Limited Scope Sustainability Assessment of Planned Nuclear Energy Systems Based on BN-1200 Fast Reactors

Introduction

Objective

This publication provides an example of the limited scope INPRO sustainability assessment of an innovative nuclear energy system using the fast reactor BN-1200 as a case study. The INPRO assessment performed at the full depth criteria level helped to identify actions, including potential research development and demonstration, for sustainable long term deployment of sodium cooled fast reactors.
This publication discusses the application of the INPRO sustainability assessment method to the innovative nuclear energy system based on fast reactor BN-1200 in the areas of economics and safety of nuclear reactors. The case study is intended to verify readiness of the updated INPRO Methodology for assessment of the sodium cooled fast reactors and to develop recommendations for further improvements and updates of the INPRO assessment method.
This publication is intended for use by organizations involved in the development and deployment of the nuclear energy systems including planning, design, modification, technical support and operation for nuclear power plants. Data provided in this publication can be used in further detailed INPRO sustainability assessments of the nuclear energy systems based on BN-1200 reactors, sustainability assessments of other fast reactors and in scenario modelling studies involving fast reactors which can be carried out by the technology holders and technology users.

Scope

Limited scope INPRO sustainability assessment of sodium cooled fast reactors was performed in 2015-2019 in parallel as a series of bilateral studies between the developers of fast reactors and the IAEA in a few countries developing such reactors. Every study was conducted as a self-assessment exercise performed by the national designer experts focused on their own design and supported by the IAEA staff.
This publication presents the results of the case study of the INPRO assessment of BN-1200 reactor in the INPRO areas of economics and reactor safety. The BN-1200 assessment has been performed by the Russian Federation experts from the Institute of Physics and Power Engineering with the support provided by JSC Afrikantov OKB Mechanical Engineering. It is based on assessors’ experience and publicly available data, taking into account proprietary information concerns.
This INPRO methodology sustainability assessment study is focused on the nuclear power plants that produce primarily electricity, heat or combinations of the two. This publication does not explicitly consider economics and safety issues related to other non-electric applications (hydrogen production, desalination, etc.) or to cogeneration involving such energy products. It is expected that as more detailed information is acquired on the safety of interactions between a reactor and industrial facilities located on the same site, the INPRO criteria and the assessment studies may be modified accordingly.

Structure

This publication follows the relationship between the concept of sustainable development and INPRO methodology areas of economics and reactor safety. Section 2 provides general information on the fast reactor development programme in the Russian Federation, to set the context. Section 3 presents the INPRO sustainability assessment of BN-1200 in the area of economics. This includes an overview of the application of the INPRO methodology area of economics to the fast reactors under development, information on the improvement of economic characteristics of sodium cooled fast reactors in the Russian Federation, basic results of analysis and sustainability assessment of BN-1200 in the area of economics. Section 4 presents the INPRO sustainability assessment on the criterion level in the area of reactor safety including assessment of the design robustness, detection and interception of anticipated operational occurrences, design basis accidents, severe plant conditions, independence of levels of defence in depth, inherent safety characteristics, passive safety systems, human factors related to safety and necessary research, development and demonstration. Section 5 summarises the discussion and suggests conclusions on the performed study.

General information on the fast reactors considered in the assessment study

Liquid metal cooled fast reactors development programme in the Russian Federation involves the following commercial sodium cooled fast reactors of BN lineage:

• BN-350 prototype reactor constructed and operated (1973-1999) in Kazakhstan;
• BN-600 reactor operated since 1980 as unit 3 of the Beloyarsk NPP, Russian Federation;
• BN-800 reactor operated since 2016 as unit 4 of the Beloyarsk NPP, Russian Federation;
• BN-1200 reactor design; pilot unit is planned to be constructed at the Beloyarsk NPP site as unit 5.

Besides that, this programme includes two more types of the liquid metal cooled fast reactors:

• Lead-cooled fast reactor design concept BREST- OD-300 with on-site manufacturing of nuclear fuel and reprocessing of spent nuclear fuel;
• Lead-bismuth cooled reactor SVBR-100 (installed electrical power rating – 100 MW(e)) which is being developed under the public-private partnership framework.

The associated closed nuclear fuel cycle facilities / technologies have been developed in the Russian Federation:

• MOX fuel experimental fuel assemblies tested in BN-350 and BN-600 reactors (approx. 50 bundles with pellet fuel and approx. 30 bundles with vibro-packed fuel);
• Mining and Chemical Combine (GHK) commercial MOX fuel fabrication facility for BN-800;
• RT-1 spent fuel reprocessing facility (PUREX technology) for uranium oxide fuel from VVER-440 (water cooled reactors) and BN-600;
• Experimental technology for the nitride fuel fabrication;
• Laboratory level technology for the pyrochemical reprocessing of irradiated nuclear fuel.

This INPRO sustainability assessment focuses on BN-1200 reactor as an example case study. In the area of reactor safety, the BN-800 reactor was selected as a reference design for BN-1200 and in a few cases (e.g. references to the accrued operational experience) the BN-600 reactor data have been used. In the INPRO area of economics the BN-800 and BN-600 reactor data have been used for estimation of trends in economic characteristics of BN reactors.
General information on BN-600, -800, -1200 reactors is presented in this section for broader context of the study.

BN-600 reactor

The original plan for the development of BN-600 reactor was based on the following basic assumptions:

• BN-600 would have higher steam temperature and pressure (540°C, 140 MPa) and increased electrical power rating compared to BN-350;
• The original basic concept and layout of BN-600 reactor were originally expected to be similar to BN-350 (loop type reactor);
• The highest electrical power rating of the standard turbo-generators designed for these parameters of steam and available at that time was 200 MW(e). The electrical power rating of BN-600 was defined as 3×200 MW(e). The reactor concept used three completely independent heat transport loops in the secondary and tertiary circuits.

The BN-600 design process modified the original conceptual ideas. Hence, the BN-600 reactor constructed at Beloyarsk NPP site contains significant improvements in comparison to the original requirements.
Loop type reactor designs have been broadly used in many countries at the early stages of development of the fast reactors. However, as sodium cooled fast reactor designs grew in installed power rating a loop type layout was revealed to have a number of complicated engineering problems.
Loop type fast reactor designs used relatively long pipelines of large cross section connecting the reactor vessel and intermediate heat exchangers. High temperature and radioactive primary sodium circulating in these pipelines change temperature at changes of reactor power level and creates essential tension/stress loads on the pipes and welds. Compensation of the thermal expansions by bending of the pipelines increases the primary circuit piping length and impedes the primary coolant natural circulation which is important for reactor cooling under the postulated station blackout conditions. Besides that, this does not provide an efficient solution since the mechanical stress in bends may become very close to the yield stress.
Moreover, the main pipeline tension/stress loads compensation forces propagate to the relatively thin walled casings of the pipes, vessels and nozzles occasionally deteriorating their stress-strain characteristics. The areas of the reactor nozzles seemed to be the most vulnerable parts and belonged to the group of most challenging components to fabricate since the stresses caused by the reactor coolant parameter cycles in these locations were the most frequent and the ranges of changing values were broad. Eliminating the nozzles provides an inherent safety feature associated with this hazard and seemed to be the most effective way to increase the robustness of the reactor design. Hence, a high power rating pool type reactor design seemed to have more reliability than a loop type designs.
The electric heating system associated with the primary coolant pipelines and the systems connected to the primary circuit of a loop type reactor had to involve sophisticated and expensive components for large isolating valves and casings of the pipes that cannot be isolated in the case of a leak. This equipment aimed to rule out any dangerous decrease in the reactor sodium level in case of the coolant leakage through the failed pipeline.
More challenges occur implementing fire and radiation protection from the sodium leaks, in particular those caused by the guillotine break of the primary pipeline – an accident scenario postulated to be accounted for by the national regulatory requirements.
Pool type layout has significant advantages compared against loop type sodium cooled fast reactors. The main reactor vessel design can avoid pipeline connections under the normal operation sodium level and accommodation of the primary circuit systems structures and components within the reactor vessel essentially reduces the probability of radioactive sodium leaks to the confinement/ containment premises and makes the solution of leak-tightness problems much easier. The reactor vessel walls, bottom and support structures are designed to withstand mechanical forces caused by the weight of reactor vessel, sodium weight and the weight of reactor internals and fuel. In the pool type design, forces coming from the thermal expansion of reactor pipelines and stressing the reactor vessel nozzles do not exist.
Placing the primary circuit systems, structures and components inside of the reactor vessel reduces the surface-to-volume ratio of the radioactive sodium and the length of welds where in the case of failure a leakage may occur. In addition to the apparently positive safety effects the cost of materials and the reactor fabrication efforts reduce significantly.
In the pool type fast reactor, accommodation of the primary circuit equipment within the large volume of sodium increases the system inertia and makes the system parameters more stable. All internals are immersed in sodium and small sodium leaks through the detachable joints between the reactor internals caused by the pressure difference on different sides can be accepted. Some of the walls of systems, structures and components placed inside the reactor vessel are not required to withstand significant strain, unlike those in the loop type reactors, and can be thinner and/or easier to fabricate. Thinner walls of reactor internals further reduce thermal stresses during the reactor transients.
Elimination of primary circuit pipelines or at least effective minimization of their length essentially reduces the cost of materials used, cost of equipment manufacturing and cost of NPP construction. Sealed compartments of the loop type design primary circuit can be eliminated, reducing the associated costs of ventilation and fire-protection systems, costs of electric heating system, costs of guard casings, thermal insulation and drains. The size and costs of several other systems can be essentially reduced, e.g. the biological shielding is required only for the reactor vessel and a few remaining pipelines and systems containing radioactive sodium. The in-vessel neutron shielding installed in the pool type reactors reduces the radiation dose to the reactor vessel and internals. It also allows for in-vessel spent fuel storage.
Another important feature of the pool type reactor is the possibility of passive removal of the residual heat in emergency situations. Passive heat removal depends on natural circulation of the primary and secondary sodium achieved through complex thermal-hydraulic design considerations in the reactor, heat exchangers and steam generators. At the time of BN-600 design development the passive removal of residual heat through the main circuits (three channels) was considered to be reliable.
The deficiencies of pool type designs are mostly related to the size of reactor vessel and to the mass of in-vessel systems, structures and components. Pool type reactor vessels are normally too large to be manufactured at the fabrication facility and transported in one piece to the NPP site. In this situation the reactor vessels need to be manufactured on the site which makes this process more expensive and challenging. The size and mass of reactor can make it vulnerable to the seismic loads and may require special arrangements different from other types of reactors. Other challenges are associated, for example with the introduction of compact and reliable in vessel neutron shielding having sufficiently long lifetime, with the reliable reactor vessel support structure and with the neutron flux monitoring necessary for the NPP power control.
Several major concepts have been incorporated in the final design of BN-600 as follows:

• Pool type layout of the primary circuit. This configuration helps to simplify the design, resolve several engineering problems associated with large fast reactors, and provides the designer with conditions and tools which allow both safety improvements and cost reductions;
• New neutron absorbers. New shim rods with higher absorbing efficiency have been installed to compensate fuel burnup effects. It allows an extension of the time between refuelling and improvement of economic characteristics for the plant;
• New design of primary coolant pumps. The primary coolant pumps use a new design with a bottom hydrostatic bearing working under sodium. This design allows control of the pump speed, reduction of the pressure in the reactor vessel gas plenum and further simplification of the layout of the primary and secondary circuits;
• New design of steam generators. Once-through sectional-modular steam generators and sodium-steam reheating scheme provide robustness for the higher temperature and pressures of steam generated in BN-600. In the case of malfunctions of the heat exchanging components the sectional design of steam generators allows isolation of a given section and uninterruptable operation of others;
• Independent heat removal loops in secondary and tertiary circuits. This layout allows for higher flexibility of the operating regimes, e.g. several reactor start-up procedures were performed consecutively and separately in every loop. The reactor can operate at power levels less than 67% of full power using only two pairs of secondary and tertiary loops. Operation of reactor at any power levels with only one pair of secondary and tertiary loops is not allowed.

The BN-600 reactor’s first criticality was achieved in 1980. The power plant unit construction used general civil industrial type building construction requirements. BN-600 has three steam generators PGN-200M, three turbines of the K-200-12.8-3 type and three electric generators of the TGV-200M type. The BN-600 has a thermal power rating of 1470 MW(th) and the electrical power rating amounted 600 MW(e).
The reactor core fuel, blankets, neutron reflectors, the control and protection system including actuators, three primary coolant pumps, most of the primary coolant pipelines, six intermediate heat exchangers and associated structures and components are placed in the main reactor vessel filled with liquid sodium (primary coolant). The volume of primary sodium exceeds 800 m³. The BN-600 uses enriched uranium oxide fuel, however it was designed to generate the ‘secondary’ nuclear fuel material (plutonium isotopes) in the reactor core and blankets.
The main reactor vessel is enclosed inside the guard vessel with the gap between these two vessels chosen to keep the sodium level in the main vessel from dropping too low in the case of main reactor vessel leak. The guard vessel sits within a concrete chamber lined with a 10 mm thick steel. The top side of this chamber has a cover of an upper biological shielding.
Each secondary loop includes two intermediate heat exchangers located in the reactor vessel, a buffer tank compensating for sodium volume changes, a secondary coolant pump, pipelines and a sectional-modular steam generator. The volume of secondary sodium in every loop equals 280-300 m³.
The pressure of gas in the reactor vessel gas plenum is normally maintained at values lower than 0.2 MPa. The characteristics of the secondary circuit including the geometry of the loops have been selected in such a way that the static pressure (including gravity pressure set by the sodium level) on the secondary side of the intermediate heat exchangers exceeds the pressure on the primary side. These arrangements help to prevent potential accidental leaks of the primary sodium into the secondary circuit through the intermediate heat exchangers.
Every once-through type steam generator has an eight-section arrangement. Every section consists of the evaporation module, steam superheater module and reheater module. Each section can be disconnected from the secondary (sodium) and tertiary (steam/ water) loops when necessary.
The basic design of the BN-600 tertiary circuit is similar to fossil power plants or secondary circuits of pressurized water cooled reactors. Every loop of the tertiary circuit includes a steam/water part of the steam generator, a turbine with its auxiliary equipment, a condenser, a deaerator, three feedwater pumps with electric drives and an emergency feedwater pump.
Three independent pairs of secondary and tertiary loops in the BN-600 reactor provide for reactor cooldown during normal operation and in the case of emergency for a safety function of residual heat removal. This three-train design may create delays in the reactor maintenance processes. Hence, a special complementary cooldown system was considered for BN-600. When introduced, this system may be connected to the secondary loops through the sodium-air heat exchangers.
BN-600 operates in a base-load mode. The average load factor estimated in the reactor design documentation amounts 76% and this value corresponds to the actual performance achieved during the reactor operation once the necessary level of technology maturity had been achieved.

BN-800 reactor

The long lasting and successful operation of the BN-600 reactor preceded the design of the BN 800, which uses most of the technologies developed and mastered at the design, commissioning and operation of the BN-600. Unlike BN-600, the BN-800 reactor design objectives involved the demonstrations of BN technology competitiveness against other energy supply options and the feasibility of industrial scale implementation of the closed fuel cycle technology.
Apart from the generation of electricity and district heating the BN-800 reactor has these design objectives:

• Operation of the reactor using mixed uranium-plutonium oxide (MOX) fuel which is important part of the closed fuel cycle technology deployment;
• Preservation and continuity of knowledge, practical skills and technologies in design, construction and operation of sodium cooled fast reactors;
• Support the research development and demonstration programmes of the sodium cooled fast reactor technologies, and the development of new fuels and of the reactor core structural materials in particular;
• Testing and validation of new systems, structures and components and new computer codes.

BN-800 is an evolutionary upgrade of BN-600. The BN-800 based its normal operation systems including the sodium circuits design and sodium systems, safety systems (except new safety systems and numerous improvements), instrumentation and control systems including reactor monitoring systems on similar systems used in the BN-600. Operating and maintenance conditions, and procedures are similar to those in BN-600, and incorporate operational approaches and experience accrued over the 60 years of the national fast reactor programme.
Commissioned in 2016 (see Figure 1), the BN-800 reactor is equipped with three steam generators N-272, a turbine of the K-800-130/3000 type and an electric generator of the TZV-800-2 type. The BN-800 has a thermal power rating of 2100 MW(th) and an electrical power rating of 880 MW(e).

FIG. 1. The BN-800 power unit at the site of Beloyarsk NPP (courtesy of Rosatom).

BN-800 is a pool type reactor using a similar layout as the BN-600 reactor described above. The volume of primary sodium in BN-800 equals 1100 m³. BN-800 can use enriched uranium oxide fuel and MOX fuel.
Like the BN-600, BN-800 uses the three circuit configuration for energy transfer and conversion. Sodium is the coolant in the primary and secondary circuits and water is the coolant in the tertiary circuit.
Each circuit consists of three identical parallel loops. Each secondary loop includes two intermediate heat exchangers located in the reactor vessel, a buffer tank compensating for sodium volume changes, a secondary coolant pump, an emergency heat removal system heat exchanger, pipelines and a sectional-modular steam generator. The volume of secondary sodium in every loop amounts approx. 350 m³. The emergency heat removal system connects to every secondary loop through the sodium-sodium heat exchanger which is connected in parallel to the steam generator.
Every steam generator consists of ten sections. Every section consists of the evaporation module and the steam superheater module. BN-800 uses steam-steam reheating scheme. Each section can be disconnected from the secondary (sodium) and tertiary (steam/water) loops when necessary.
The layout of tertiary circuit and its systems structures and components have similarity to superheated steam turbines. The tertiary circuit through the primary and secondary circuits and steam generators provides the removal of heat from the reactor at normal operation conditions, deviations from normal operation and in most accident scenarios. However, this heat removal scheme may fail in the cases of loss of feedwater, station blackout scenarios, earthquakes, etc. Unlike BN-600, the BN-800 uses a single feedwater supply system for all three steam generators which stipulates an emergency heat removal system connection in parallel to the steam generators.
The most extensive and sophisticated improvements of the BN-800 design were introduced in the systems, structures and components important to safety, and mostly incorporating new regulatory requirements. The design of BN-800 has been developed to be in compliance with the new or updated national regulatory documents on nuclear safety such as OPB-88/97, PBYa RU AS-07, SP-AS-03, etc. Classification, layout, design and construction of the NPP buildings and premises in accordance with the fire and explosion protection requirements complied with Russian fire safety standards norms documented in NPB 105-95. Nevertheless, all systems modifications and functions included the operational experience of previous installations.
BN-800 protective safety systems include:

• Emergency heat removal system;
• Reactor protection system;
• Reactor loss-of-coolant protection system;
• Reactor overpressure protection system;
• Secondary circuit overpressure protection system;
• Heat removal system for fuel assemblies for use during their reloading from the reactor to the spent fuel ‘drum’ storage;
• Spent fuel ‘drum’ storage heat removal system;
• Spent fuel ‘drum’ storage casing overpressure protection system;
• Guard casings of primary pressure gas pipelines and guard shell of the pressure chamber.

BN-800 confinement safety systems include:

• Reactor guard vessel;
• Reactor core catcher;
• Reactor confinement compartments and leak-tight enclosure;
• Guard casings of primary auxiliary systems pipelines;
• Primary sodium systems, structures and components compartments ventilation system and spent fuel ‘drum’ storage compartment ventilation system;
• Sodium systems, structures and components compartments fire protection system;
• Spent fuel ‘drum’ storage casing;
• Guard casings at the pipeline sections from the spent fuel ‘drum’ storage to the overflow vessel;
• Exterior lining of the spent fuel cooling pond.

In comparison with the BN-600, the modifications of the BN-800 reactor design include the following features increasing the reactor systems reliability and improving safety:

• Sealed cover was introduced around the reactor pressure chamber;
• A reactor vessel bottom part and the reactor support structure have been redesigned to improve seismic characteristics;
• The thickness walls of the reactor main vessel and guard vessel was increased from 20 to 30 mm;
• Reactor core catcher has been introduced to protect the reactor vessel from the molten fuel effects in the case of severe accident;
• Control rods with the passive actuation principles have been added to the reactor protection system;
• The primary coolant purification system has a stationary arrangement installed for separation and removal of caesium from sodium;
• System protecting the reactor from overpressure or accidental depressurization has been improved;
• New ionization chambers have been introduced to monitor the reactor core in the subcritical state;
• New reactor core design was proposed to minimize the value of sodium void reactivity effect;
• One rotating plug was added to the fuel reloading system; however, the reloading system and reloading procedures have been simplified and one in-vessel elevating machine was eliminated;
• In-vessel fuel cladding leaks detection system has been installed and interlocked with the reloading system;
• Entrainment of primary sodium from the reactor vessel to the systems and pipelines located outside of reactor vessel in the case of depressurization was ruled out;
• Fire and explosion protection of the steam generators and protection of other systems, structures and components located in the steam generator compartments have been improved;
• New emergency heat removal system has been installed and connected to the secondary circuit loops in parallel to the steam generators. The residual heat removal is provided through the independent sodium-air heat exchangers;
• Unlike the BN-600 electric generators TGV-200M that uses hydrogen for cooling, the cooling system of the BN-800 electric generator TZV-800-2 uses water as coolant;
• Filters removing radioactive aerosols from the combustion products have been added to the fire protection and ventilation systems.

BN-1200 reactor

The Russian Federation nuclear power strategy in the first half of 21 century determines tasks for the new generation nuclear power plant development and deployment:

• Eliminate accidents that require public evacuation or relocation;
• Use effectively fissile and fertile materials provided by natural uranium;
• Multi-recycling of nuclear material minimizing amount of high level radioactive waste;
• Strengthening of non-proliferation characteristics of materials and technologies;
• Maintain competitiveness of nuclear power.

Minimum excess reactivity in the reactor core can be achieved through the appropriate breeding characteristics. Effective use of natural uranium resources can be achieved using a closed fuel cycle and renouncing of direct disposal of the spent fuel. Multiple recycling of nuclear materials in closed fuel cycle can help to minimize the minor actinide content in the high level waste.
The development of large sodium cooled fast reactors started in the Russian Federation a few decades ago. Design work on the BN-1200 reactor started in 2007 as part of the JSC Concern Rosenergoatom programme. Approval of the terms of reference for the development of BN-1200 reactor installation occurred in 2012. Later the development of BN-1200 moved ahead as part of the national ‘Breakthrough’ programme.
By the time of this INPRO assessment study, the basic design of the reactor installation and systems, structures and components had been developed including the design of steam generators and the core and fuel design of both MOX fuel and high density (nitride) fuel. The development of BN-1200 detailed design has been a continuous project.
The ‘Breakthrough’ Project Office and the national nuclear utility JSC Concern Rosenergoatom coordinate the BN-1200 project. The ‘Breakthrough’ Technical Committee reviews major modifications, new features and other technical issues related to the reactor development. JSC OKBM is the BN-1200 design organization responsible for the implementation of RD&D programme and design developments required. Organizations participating in the development of systems and technologies for BN-1200 have many years of experience in this area and established cooperation mechanisms, which are necessary for the assurance of design quality and safety.
The ‘new generation’ reactor designs which are being developed in the Russian Federation, including BN-1200, aim to resolve the following tasks:

• Competitiveness with other advanced nuclear power plants and with power plants using fossil fuel;
• Enhanced safety level corresponding to the requirements formulated for the Gen IV reactors, and elimination of the necessity of public evacuation/relocation in the case of potential accidents;
• Multi-recycling of plutonium isotopes in fast reactors and nuclear fissile material breeding for long term fuel supply to other types of reactors (water cooled reactors);
• Duration of the construction period up to 48 months for ‘N-th-of-a-kind’ commercial power plants;
• A feasibility of commissioning of a series of NOAK commercial power plants in 2-3 years after the FOAK power plant commissioning.

The following basic conceptual provisions have been defined as a basis for the BN-1200 design development project:

• Proven technologies and experience acquired in the design, commissioning and operation of BN-600 and BN-800 reactors should have broad use in the BN-1200 design to the greatest extent possible ;
• Testing and validation of improvements in the reactor safety, economic competitiveness and effectiveness of fuel management incorporating innovative technologies through the appropriate RD&D activities using existing and newly developed research facilities;
• Optimization of the infrastructure requirements can be achieved through the selection of the appropriate value of the BN-1200 electrical power rating and may involve unification of requirements for the NPP siting and unification of the electric generators and electric components used for the plant connection to the grid;
• Transportation of the NPP components to the construction site by railroad.

BN-1200 is a pool type reactor and it has an integral layout of the primary circuit like BN-800 and BN-600 where the primary circuit and radioactive primary coolant are in the main reactor vessel enclosed in the guard vessel as described above.
Like the BN-600 and BN-800, the BN-1200 reactor vessel is supported at the lower cylindrical part from the bottom side. Cooling of the reactor vessel has been worked out in a way that shielding structures provide the possibility of compact layout of the in-vessel systems, structures and components, and allows to get relatively small vessel diameter.

FIG. 2. BN-800 and BN-1200 reactor vessels

Unlike the BN-600 and BN-800, the BN-1200 reactor maintains the level of primary sodium below the tapered upper part of the main vessel. This modification eliminates the need of guard vessel and bellows in the tapered part of the reactor vessel. Use of the main reactor vessel upper tapered part for the support of reactor equipment allows accommodating the four primary coolant pumps, four intermediate heat exchangers, the emergency heat removal system heat exchangers and a cold filter trap of the primary circuit within the BN-1200 reactor vessel without increasing the vessel diameter. The latter modification eliminates components containing primary sodium outside of the reactor vessel and eliminates potential leaks of primary sodium from the reactor vessel external system. Figure 2 summarises basic similarities and distinctions between BN-1200 and BN-800 reactor vessel designs.
The four in-vessel ionization chambers improve the BN-1200 neutron flux monitoring and eliminates the need of neutron guides used in BN-600 and BN-800. A rotating roof planned for installation above the BN-1200 reactor vessel protects the reactor systems, structures and components from the potential falling of heavy objects.
Like BN-600 and BN-800, the BN-1200 uses the three-circuit configuration for the energy transfer and conversion. Sodium is used as a coolant in primary and secondary circuits, and water in tertiary circuit. However, in the BN-1200 every circuit contains four loops.
Each secondary loop is physically separated from three other secondary loops and located in a separate compartment. The loop includes a single intermediate heat exchanger located in the reactor vessel, a secondary coolant pump, pipelines and a sectional-modular steam generator. In BN-1200 the guard casings is introduced in the most of secondary pipelines. The buffer tanks compensating for the secondary sodium volume changes is combined with the tanks of secondary coolant pumps.
In order to reduce the length of pipelines in the secondary circuit and to minimize the number of installed isolating valves the bellows, compensators have been introduced in the BN-1200 reactor. Further reducing the volume of building and materials used the shell-and-tube type steam generators have been introduced in BN-1200 instead of sectional-modular type steam generators in BN-600 and BN-800 (see Table 1 and Figure 3).

Table 1. Volume of the fast reactor buildings

Specific volume	BN-800	BN-1200
Main building, m3/MW(e)	750	525
Steam generators boxes, m3/MW(e)	32	16

FIG. 3. BN-800 and BN-1200 reactor building schemes.

The tertiary circuit layout of BN-1200 includes the K-1200-16.0/50 turbine and steam-steam reheating scheme. To increase the plant thermal efficiency to 43.6% (gross), the temperature and pressure of live steam, and feed water have been increased compared to the BN-800 parameters. BN-1200 uses electric generator of the type which is used in other advanced water cooled NPPs of that power rating designed in the Russian Federation (AES-2006).
In comparison with BN-600 and BN-800, other than the discussed above modifications of BN 1200 reactor design for increasing the reactor systems reliability and improving safety include the following:

• Use of uranium-plutonium nitride fuel helps to reduce the excess reactivity at full power conditions mitigating the potential consequences of RIA type of accidents. The BN-1200 excess reactivity at full power conditions will not exceed 0.5% Δk/k;
• Simultaneous withdrawal of more than one control rods from the reactor is prevented by multiple independent measures;
• Passive high temperature actuated control rods system is introduced in BN-1200 in addition to the passive hydraulically suspended control rods system20 and active shut down system;
• Maximum power density in the reactor core is reduced to 380 MW/m³;
• Duration of spent fuel storage within the reactor vessel is extended to two years reducing the spent fuel assembly power density to 2 W/cm³, simplifying the fuel reloading process and improving safety of spent fuel management;
• Emergency heat removal system is based on the passive removal of heat from the reactor to the environment through the sodium-sodium and sodium-air heat exchangers, and designed in such a way that in the case of emergency cooling the primary sodium circulates through the reactor core fuel assemblies driven by natural circulation flow;
• The above the reactor premise is used for confinement of radioactive gases and aerosols prior to subsequent filtration in the special ventilation system reducing the radioactivity of released gases. Calculated core damage frequency of the BN-1200 reactor (approx. 5×10^-7 1/a) is essentially reduced compared with those of BN-800 (approx. 2×10^-6 1/a) and BN-600 (approx. 10^-5 1/a).

INPRO sustainability assessment of BN-1200 in the area of economics

This section presents an overview of the assessment method used in the INPRO area of economics and modifications to the method that have been introduced in the study. It provides input parameters used in the calculations and basic assumptions made by the national experts performing such calculations. Finally, the results of calculations presented in this section may be considered as an input for further research activities focused on optimization of the fast reactors designs and fuel cycles, however they should not be considered in relation to any commercial activities.

Overview of the application of the INPRO methodology area of economics to the fast reactors under development

The INPRO methodology has been developed for the sustainability assessment of the nuclear energy systems comprising different types of reactors and fuel cycle facilities. Sustainability assessment using INPRO methodology covers several different areas including economics, reactor safety, safety of fuel cycle facilities, waste management, etc. However, in the area of economics the INPRO methodology focuses on economic characteristics of energy products, i.e. electricity, heat, etc, generated by an NPP, rather than on economic parameters of fuel cycle facilities which are used as an input in a form of ‘cost of a given service’. For example, unless such calculations are the only way of obtaining costs of services the assessor is not required to calculate the figures of merit of the enrichment or reprocessing facility for the purpose of INPRO sustainability assessment. Instead, the costs of enrichment or reprocessing services are aggregated along with other costs in the figures of merit of the NPP producing final energy product. These figures of merit are used for the assessment against INPRO criteria in the area of economics.
INPRO methodology recommendations have the same hierarchy in all areas of assessment. On top of this structure a basic principle of sustainability is defined for every area determining a goal to be achieved to make the nuclear energy system sustainable. The necessary actions to be taken to achieve the goal defined in the basic principle are introduced on the second level and called ‘user requirements’. At the next and third level, every user requirement splits into a few criteria. Criteria are the tools for assessment. Every criterion consists of an indicator (e.g. parameter) and an acceptance limit defining the range of values satisfying this criterion.
The INPRO assessment is normally done on the criteria level. When all criteria comprising a given user requirement are met, it means that necessary action occurred in a correct manner to meet the user requirement. When all given user requirements in a given area are met, the goal defined in the corresponding INPRO basic principle is also met. However, the INPRO methodology assessment is not only used for the confirmation of nuclear energy system sustainability. The INPRO assessment criteria which are not met provide valuable information on the gaps in the nuclear power programme and help to define follow-up actions (e.g. R&D) necessary to close these gaps. Comparison of the different assessment results may help to estimate the potential advantages of the assessed nuclear energy systems.
Judgement on the system sustainability can be derived from the results of full scope INPRO assessment in all INPRO methodology areas. However, preliminary results of the assessment in selected areas can be used for the optimization of nuclear energy system or modification of the selected system components. Although such limited scope assessments can provide valuable information on the system sustainability, the judgement on sustainability can be concluded only in the full scope INPRO sustainability assessment study in all areas.
This limited scope assessment study omits INPRO areas involving the essential numbers of country specific criteria (waste management, infrastructure, proliferation resistance) and focuses on the two areas encompassing the most of design related information – economics and reactor safety.
The INPRO methodology in the area of economics sets up the goal of achieving ‘affordable and available’ produced energy and related products and services in a sustainable nuclear energy system. The affordability of energy is understood as cost competitiveness with alternatives available in the country or region. The availability of nuclear energy is seen as the ability to finance the project at acceptably low investment risk.
Four user requirements have been introduced in the INPRO methodology area of economics to specify how the basic principle can be met:

• The cost of energy supplied by nuclear energy systems, taking all relevant costs and credits into account, should be competitive with that of alternative energy sources, that are available for a given application in the same time frame and geographic region/jurisdiction;
• The total investment required to design, construct, and commission nuclear energy systems, including interest during construction, should be such that the necessary investment funds can be raised;
• The risk of investment in nuclear energy systems should be acceptable to investors;
• Innovative nuclear energy systems should be compatible with meeting the requirements of different markets.

The developer of sustainable energy technology is expected to meet these user requirements. The INPRO assessor has eight criteria which can be used for checking the status of nuclear energy system in relation to the INPRO user requirements:

• Cost competitiveness;
• Attractiveness of investment;
• Investment limit;
• Maturity of design;
• Construction schedule;
• Uncertainty of economic input parameters;
• Political environment;
• Flexibility.

Only five of these eight criteria have been assessed in this study as explained below. The cost competitiveness criterion uses the cost of energy generated by the nuclear energy system as an indicator, i.e. characteristic to be assessed. The limit for acceptance of this characteristic was defined as:
$C_{N}\leq k\times C_{A}$ (1)
where 𝐶_N is cost of nuclear energy, 𝐶_A is cost of energy from alternative source, and k is a factor based on strategic considerations.
The INPRO methodology uses the levelized unit energy cost (LUEC) concept for definition of the energy cost (both 𝐶_N and 𝐶_A). LUEC of an assessed NPP should be lower or at least comparable, within a factor of k, to the LUEC of a competing power plant type. LUEC of an NPP and of a given alternative power plant can be calculated using the NEST tool developed by the IAEA or other tools using similar approach for the calculation of economic functions of NPPs and alternatives. LUEC comprises three major constituents: capital costs, operation and maintenance costs and fuel costs. It was introduced as the equivalent of the electricity price at zero profit covering all the capital, operating and maintenance and fuel costs with regard for the discount rate.
LUEC calculations have to include contingency and all accountable external costs at a discount rate associated with specific investment policy. Accountable external costs may include the costs of radioactive waste management, decommissioning or back-fitting costs for the NPP, greenhouse gas emission fees and charges for fossil or biomass fuelled power plants, cost of power banks or backup power plants for wind and photovoltaic power plants.
The assessment of NPP deployment differs from the assessment of NPP design under development. In the case of NPP deployment, the alternative power plant is normally selected among non-nuclear power plants available in the country and, if applicable, among NPPs operating there. In the case of development of a new NPP design the competing power plant could be a comparable NPP of different existing or developing design. Nevertheless, LUEC needs to be calculated for the ‘N-th-of-a-kind’ (NOAK) case (the ‘first-of-a-kind’, FOAK, case can be evaluated separately). In this study VVER-TOI (new water-cooled thermal reactor under development) and BN-800 (newest operating sodium cooled fast reactor) have been selected as alternative power plants.
Competitiveness does not mean that the assessed nuclear energy system has to produce the cheapest energy in the country or region. The strategic considerations factor k in Eq.(1) depends on particular circumstances in a given territory. It was introduced to account for factors which have not been evaluated in terms of costs and not included into the energy costs 𝐶_N and 𝐶_A, for example: energy supply security, energy costs stability, diversity of energy supply, unaccounted environmental impacts, utilization of domestic resources (mineral or labour) and industrial capacity, public and political support, etc. Such considerations may be used as justification for k factor values different from 1.
Since economics does not provide an aggregated universal function for the complete economic assessment of the energy system options and LUEC does not present a full economic picture of a given project the next INPRO criterion, attractiveness of investment requires to calculate and assess the economic (financial) figures of merit. There are three financial figures of merit recommended for the INPRO assessment:

• Internal rate of return (IRR);
• Return on investment (ROI);
• Net present value (NPV).

The limits for acceptance of these characteristics are defined as:
$IRR_{N}\geq IRR_{A}$(2)
$ROI_{N}\geq ROI_{A}$ (3)
$NPV_{N}\geq NPV_{A}$ (4)
where IRR, ROI and NPV are internal rate of return, return on investment and net present value of the assessed NPP, and IRR, ROI and NPV are internal rate of return, return on investment and net present value of the alternative power plant (alternative NPP in the case of this study). Since net present value represents the total net value of the investment, discounted to time 0, its absolute value will depend on the size of the investment and it needs to be normalized to the initial (discounted) capital investment made up to the start of plant operation or to the power output of the plant. When comparing the economic functions of assessed NPP with the alternative energy supply options using Eq.(1)–Eq.(4), if energy planning has identified nuclear power as part of an optimized mix of generating options the comparison of C_N, IRR_N, ROI_N, NPV_N, C_A, IRR_A, ROI_A, NPV_A, is not, of itself a determining consideration. But if limits defined in Eq.(1)– Eq.(4) are not met, the assessor needs to set out the explanations of why the difference is acceptable in the circumstances. The next INPRO criterion in the area of economics, investment limit, requires that the highest single plant total investment up to the commissioning of the reactor be compatible with the ability to raise capital in a given market climate. This criterion is relevant to the situations when the construction of NPP is financed by foreign or private investors. The assessment of this criterion involves calculation of two parameters:

• Total investment which consists of the overnight capital, the interest during construction, contingency allowances, owner’s cost, back fitting and decommissioning costs;
• Investment limit – the maximum level of capital that could be raised by a potential investor in the market climate.

The calculation of the second parameter is based on a country specific set of input data. The INPRO assessment of BN-1200 is focused only on the domestic deployment of these reactors which assumes that necessary investments will not be provided by foreign or by private investors. Therefore, the assessment of criterion ‘investment limit’ was omitted in this study and necessary clarification was added.
To assess the risk of investment in the nuclear energy system the INPRO methodology provides a group of four generic criteria focused on different factors that can theoretically impinge the NPP project. The first criterion in this group, maturity of design, is focused on the evaluation of technical development status and licensing status of the design. At the time of development of this report (2019) the licensing process for BN-1200 has not been started yet and this criterion ‘maturity of design’ has not been assessed in the INPRO area of economics. It can be assessed at the later stages of the BN-1200 programme.
The next criterion, construction schedule, considers the background information used for the definition of times necessary for the reactor construction and commissioning. These times are among the most sensitive parameters for the energy cost calculations and their realistic definition is important for the investment risk minimization. Information on simplification of the BN-1200 design and improvement of the construction methods is discussed in this report section on the INPRO assessment of BN-1200 in the area of reactor safety. The construction period of the BN-1200 reactor is assumed to amount 6 years. However, due to the lack of practical experience on BN-1200 construction and commissioning, the criterion of ‘construction schedule’ has not been assessed in the INPRO area of economics. It can be assessed at the later steps of BN-1200 programme.
The assessment of uncertainty of economic input parameters requires an analysis of the sensitivity of important input parameters. Sensitivity analysis can involve many different methods and approaches, e.g. calculation of robustness indexes, Monte-Carlo studies, cost sensitivity diagrams, etc. In this study, the sensitivity analysis of BN-1200 economic parameters used cost sensitivity diagrams mainly for screening purposes. At the next stages of project development this analysis can be expanded to other methods.
The last criterion assessing the risk of investments, and political environment, requires checking the long term commitment to nuclear energy system development/deployment in the country.
The last criterion of the INPRO methodology area of economics, flexibility, evaluates the potential of the reactor to meet different market conditions including both electricity markets and fuel markets. Electricity market conditions may involve such characteristics as the grid size and requirements on participation in the grid regulation. Fuel market considerations include the possibility to use different types of fuel (e.g. UOX, MOX, nitride or metallic fuels) or fuels fabricated by different suppliers without major modification of the installation. To meet this criterion an NPP is expected to be sufficiently flexible to provide competitive energy for a wide range of markets.

Improvement of economics characteristics of sodium cooled fast reactors

Strategic planning of the nuclear energy system which is expected to contribute to the sustainable development in a global prospective or in a defined country/region involves several stages (e.g. system modelling, assessment, definition and implementation of the follow-up measures) and iterations at different levels of the nuclear energy system maturity. Information on the process of nuclear energy system development and optimization, on the requirements, boundary conditions, continuity of support and the trends related to improvement of the system economics can provide valuable background for the evaluation of potential investment risks, justification of basic assumptions used in cost calculations and input data reliability.
The fast reactor programme implemented in the Soviet Union and later in the Russian Federation has demonstrated scope and continuity (Figure 4). Passing from one stage of the programme to another, accumulating necessary experience and gradually improving the technology avoided overhasty decisions in and minimized potential risks from the introduction of innovative technology. R&D studies of the advanced systems and optimization of the reactor design are going on continuously. New reactors are characterized by improved operating parameters, higher fuel burnups, and improved safety. Prospective BN reactors with dense fuels allow an increase in the breeding of fissile material (total breading ratio and breading ratio in the core reaching the values of 1.45 and ~1 respectively). Different schemes of recycling minor actinides are under investigation with the objective to reduce the amount and radiotoxicity of HLW.

FIG. 4. Fast reactors developed in the Soviet Union and in the Russian Federation.

The fast reactor programme in the Soviet Union was launched in 1950s, when IPPE commenced the development of experimental fast reactors. In 1956-57 the design of sodium cooled fast reactor BR-5 was developed. This reactor commissioned in 1958-59 in IPPE originally had thermal power rating of 5 MW(th). Moving on into the 1960s, IPPE performed a comprehensive comparative analysis of different coolants and defined a preference for a fast reactor technology concept based on the sodium cooled fast reactor with steam-turbine cycle for the energy conversion.
The second Russian sodium-cooled fast reactor, BOR-60, was commissioned in 1969. For a long time, it was used as the main experimental facility to study sodium-cooled fast reactors. The convenience of BOR-60 design for conducting of a variety of research studies allowed to use this experimental facility extensively in the development and justification of the first Soviet commercial sodium-cooled power reactor BN-350.
The Soviet Union constructed the prototype commercial fast reactor BN-350 in early 1970s with its commissioning in 1973. The BN-350 was a loop-type reactor, having a designed lifetime of 20 years. It performed both electricity production and water desalination over its 25 year operating period (the BN-350 was sited in western Kazakhstan). From the beginning it provided valuable information on the real scale systems, structures and components behaviour which assisted the development of the BN-600 reactor constructed several years later.
The BN-600, commissioned in 1982, a commercial pool-type pilot reactor with an electrical power rating of 600 MW(e) demonstrated the results of significant design improvements and economic optimization. Like its predecessor BN-350, the BN-600 reactor uses enriched uranium oxide fuel. Over the next 30 years of operation its load factor achieved ~ 76% which is close to the load factors of traditional water cooled reactors. The average number of unplanned total scrams per 7000 hours critical during the period from 1990 to 2009 was 0.2 . Further improvement of the technology and economic optimization was undertaken during the BN-600 operation. It involved the update or revision of the maintenance and replacement techniques for selected systems, structures and components, including the major power plant components such as pumps, heat exchangers and steam generators. Valuable experience was accrued on prevention and mitigation of potential sodium leaks, demonstrating effectiveness of the inspection methods, monitoring and confinement systems, and fire protection systems. More information on the BN-600 characteristics is provided in Section 2. The newest operating reactor in the BN lineage, BN-800, was connected to the grid in 2016. The BN-800 design concept was developed in the period from 1983 to 1993 incorporating lessons learned from the successful operation of BN-600, lessons from the Chernobyl accident and the revised regulatory requirements introduced in the Russian Federation. Revised regulations required ensuring the safety of local population during the design basis accidents without such protective measures as evacuation or relocation. The BN-800 design was the first Russian reactor that obtained a construction license from the Federal Nuclear and Radiation Safety Authority after the Chernobyl accident. Unlike its predecessors, the BN-800 reactor was designed to be operated with MOX fuel and in a few years after commissioning it has been using both UOX and MOX fuel assemblies in the core. This reactor is expected to play important role in the development and refinement of closed fuel cycle technologies in the Russian Federation. The objective is to obtain the MOX fuel burnup of 15% heavy atoms and higher, and to test fuel rods and fuel assemblies with the nitride fuel having a higher density than MOX. Improvements of the spent MOX-fuel reprocessing and the recycled fuel re-fabrication technologies are being carried out in parallel. The BN-800 will be used for the development of technology to burn minor actinides accumulated in the spent fuel from different types of reactors (fast and thermal). More information on the BN-800 characteristics is provided in Section 2. The next step of development of the BN reactor family is the large commercial reactor BN-1200. The concept of BN-1200 envisages reactor core operation using different types of fuel and permits variation of fuel utilization parameters in accordance with system requirements providing significant flexibility and possibility to adapt the reactor to different market conditions. The optimization of the BN-1200 reactor design involved multiple modifications of the systems, structures and components layout, providing significant improvement of the reactor economics characteristics. Most of BN-1200 performance characteristics are similar to the characteristics of traditional large water cooled reactors which are currently under development in the Russian Federation. The installed power rating of BN-1200 reactor (1220 MW(e)) is similar to that of VVER-TOI (1255 MW(e)) and the design lifetime of both reactors is 60 years. Specific (per MW(e) installed) capital costs of the BN-1200 reactor are less than half of those for BN-350 and achieved the level of VVER-TOI. The reduction of mass of metal used in BN-1200 (per MW(e) installed) compared against BN-350 reactor is even lower. The evaluation of the improvement of the two latter characteristics is presented in Figure 5.

FIG. 5. Specific mass of metal used and specific capital costs for different BN reactors at different stages of the BN technology development programme.

The BN-1200 reactor core is designed for using either MOX fuel similar to BN-800 or a new type of fuel with a higher density and plutonium and uranium in nitride form. As in BN-800, the BN-1200 fuel assemblies are designed in such a way that sodium plenum is maintained in the upper part of the core. However, the increased fraction of fuel per unit of the core volume yields the breeding ratio of about ~1.2 and maximum fuel burnup of at least 15% heavy atoms.
At the time of this report preparation the optimization of BN-1200 design has not been fully completed and further improvement of the economics characteristics can be reasonably expected. However, the general trends used in this study and presented in this report remain valid and input information is sufficient for the limited scope INPRO assessment.

Basic results of analysis in the area of economics

This section presents the results of limited scope analysis of the planned system based on the sodium-cooled fast reactor BN-1200 in the Russian Federation.
Preliminary economic studies of an energy system development, deployment or modification involve several steps including system planning and modelling, cost study, profit characteristics or cashflow analysis, sensitivity study, etc, evaluating the system viability, competitiveness, and attractiveness which can be further studied in the comprehensive and somewhat cumbersome stages of financial analysis. Both the system planning and modelling study and the financial analysis are outside of the scope of the INPRO methodology sustainability assessment. The energy system planning and modelling and, more specifically, the nuclear energy system planning and modelling are the necessary prerequisites of the INPRO assessment study.
In this INPRO assessment of the BN-1200 reactor there is the assumption that the necessary scenario studies have been successfully performed, involving the fissile/fertile materials flow analysis, and the role of the nuclear energy system is understood. INPRO sustainability assessment in the area of economics consists of the consideration of the energy cost study results, profit characteristics and sensitivity study.

Cost of energy generation

Overnight capital cost

Projections of overnight capital costs associated with a given NPP design are sensitive to the assumption on the system characteristics at the different stages of technology maturity. Overnight costs of a FOAK reactors are different from NOAK reactors. The latter costs may depend on the assumed number of commercial reactors to be deployed in the home country and abroad. Prototype reactors normally have lower capacities than commercial designs. If the difference is too large the estimation of the effects of the economy of scale reducing the specific cost of energy, may introduce essential uncertainty. Smaller difference in the installed capacities allows better estimation of costs.
Another important assumption is related to the number of power units at a given site. Due to the common use of several systems, structures and components by all power units at the site, depending on the reactor design, the cost of a second power unit of the same design may be reduced by a factor of approx. 1.5 (estimated by the IPPE). The cost reduction effects from construction of additional units are essentially smaller and can be estimated at less than 5%.
Combinations of these assumptions may influence both the overnight costs of an NPP, and the cost of energy generated. For example, the overnight cost of a single 1 GW reactor can be lower than the capital costs of two 0.5 GW alternative units located at different sites, however, depending on the effects of economy of scale, it can be higher than the overnight costs of two 0.5 GW alternative units located at one site. The comparison of the two 1 GW reactors located at one site against alternative power plants of the same total capacity may yield different results.
Cost calculations in this publication have been performed for the case of twin unit NPPs. However, these calculations do not account for the scenario of reactors deployment, i.e. the difference in dates of the twin reactors’ commissioning and discounting of the costs between these dates has not been accounted for. Average energy costs have been calculated assuming that twin units had been constructed and commissioned simultaneously.
Overnight costs of the twin unit NPPs with BN-800 reactors, BN-1200 reactors and VVER TOI reactors are presented in Table 2. A FOAK BN-1200 is planned to be constructed at the site of Beloyarsk NPP and the estimation of overnight cost was performed for that site. Note that there is no plan to deploy BN-800 reactors in the Russian Federation and the overnight cost of the twin BN-800 NPP was estimated numerically to make this comparison more convenient. Estimation of the overnight costs of VVER-TOI has not been made related to any specific site and potential site specific requirements.

Table 2. Estimated overnight capital costs as of 01.01.2013

Reactor type	Overnight cost per twin unit power plant, 109 USD	Speicific overnight cost, 109 USD/kW(e)
BN-800 (2 х 880 MW(e))	6.77	3.8
BN-1200 (2 х 1220 MW(e))	8.22(FOAK)	3.4(FOAK)
	7.86(NOAK)	3.2(NOAK)
VVER-TOI (2 х 1255 MW(e))	7.72	3.1

Overnight capital costs of the FOAK BN-1200 are estimated at approx. 10% lower than the capital costs of BN-800 from the economy of scale and optimization of design surpassing the cost of BN-1200 safety improvements. Further optimization of BN-1200 design is expected to yield approx. 5% reduction of the NOAK BN-1200 overnight costs achieving the level of overnight costs of the new Russian water cooled reactors VVER-TOI.

Operation and maintenance cost

The operation and maintenance costs usually consist of the following expenses:

• NPP personnel salaries;
• Cost of services necessary for NPP operation provided by the external contractors including repair, maintenance, inspections, safety assessments, meteorological and environmental studies, etc;
• Cost of electricity, fossil fuels, chemical materials, etc, necessary for NPP operation/maintenance;
• Cost of equipment necessary for NPP operation/maintenance including cost of the NPP components which the NPP design calculates as necessary for occasional replacement;
• Cost of licensing related activities and services (peer reviews, assessments, inspections, knowledge management and personnel training, etc);
• Retrofit costs during the NPP operation;
• Cost of management of radioactive waste other than spent fuel or waste from spent fuel reprocessing;
• Decommissioning and backfitting costs;
• Insurance fees, cost of financial services, etc.

Operating and maintenance costs do not include the amortization costs, fresh fuel costs, cost of spent nuclear fuel management outside of NPP, cost of spent fuel reprocessing and management of radioactive waste arising from reprocessing. One part of operation and maintenance costs (e.g. NPP personnel salaries) does not depend on the amount of energy generated by NPP and needs payment on a regular basis regardless of operational mode. Remaining operations and maintenance costs are variable and depend on the average amount of energy generated by an NPP. However, those dependencies can be different.
At this stage of the BN-1200 development the detailed characteristics of the constituent pieces of operation and maintenance cost were not available to the assessor. For the purpose of INPRO sustainability assessment all operation and maintenance costs have been considered as fixed values independent from the NPP average load factors. The operation and maintenance costs of BN-800, BN-1200 and VVER-TOI are presented in Table 3.

Table 3. Operation and maintenance costs

Reactor type	Annual specific operation and maintenance costs, USD/kW(e) per annum
BN-800 (2 х 880 MW(e))	134
BN-1200 (2 х 1220 MW(e))	122
VVER-TOI (2 х 1255 MW(e))	102

In this study the operation and maintenance costs of new fast reactors were estimated using the available data from water cooled reactors and the results of analysis of BN reactors design characteristics and operating/maintenance procedures. For example, BN-1200 personnel salaries are estimated as proportional to the number of employees per unit of installed power rating with a surcharge for MOX or nitride fuel management.

Fuel costs

Fuel cost relative contribution to the overall cost of energy generated in an NPP is relatively modest. It varies depending on the type of reactor and fuel cycle, national policies, company strategy, selected investor and vendors, however it normally remains within one fifth of the total energy cost. Theoretically, the type of fuel cycle can affect the sustainability areas other than economics, e.g. waste management or environment, and the fuel cycle consideration can involve issues other than costs. The closed fuel cycle fuel cost calculations can be quite sophisticated. However, in many cases their primary objective is rather to demonstrate that increased complexity of the process does not make the overall cost of fuel unacceptably high.
NPP fuel cost is the aggregated value of expenses born at different steps of the complete fuel cycle, both frontend and backend, including the final disposal of spent fuel in the case of once-through fuel cycle or the disposal of high level waste in the case of using reprocessing. The frontend cost of once-through fuel cycle (LEU) is combined from the cost of natural uranium and costs of fuel cycle services necessary for obtaining the form of fuel which can be safely used in a reactor type. The backend involves costs of services necessary for obtaining the safe end state of spent fuel (deep geological disposal). The costs of fuel cycle services strongly depend on the scale of fuel cycle facility providing a given service (economy of scale) and the scale of facility depend on the demand of this service, i.e. on the scale of nuclear power programme.
In the case of closed fuel cycle the backend services separate fissile and fertile materials from the waste, move waste to the end state and feed the frontend with fuel materials. Apparently, this feedback implies that the costs of backend services, e.g. cost of spent fuel reprocessing, and the reactor fissile material breeding characteristics affect the cost of ‘fresh’ fuel and can introduce additional uncertainty to the fuel cost calculation. A few other process step links need to be accounted for, e.g. better refining of fissile material may increase the cost of spent fuel reprocessing, however it can reduce the cost of other steps of the closed fuel cycle. The cost evaluations considering different types of the reactors and fuel cycles in a nuclear energy system may involve more links . The evaluation of ranges of the costs of different fuel cycle services is provided in Ref.. The ranges are quite broad, for example, the cost of MOX fuel fabrication varies from 1000 USD/kg to 6000 USD/kg and the cost of MOX fuel reprocessing – from 640 USD/kg to 2600 USD/kg. The discrepancies in cost estimations may be related to the lack of experience, different assumptions and criteria.
Data on the dependence of fuel cycle services costs from the production scale of a fuel cycle facility are relatively scarce and mostly relate to the once-through fuel cycle and uranium fuel production. Refsprovide evaluations for the fuel fabrication facility producing MOX fuel for the light water reactors. Figure 6 presents cost vs production rate diagram for the light water reactor MOX fuel fabrication facility.

FIG. 6. MOX fuel fabrication cost as a function of facility production rate (modified).)

Higher facility production rates allow fabricating MOX fuel at lower cost. Depending on the production rate the cost of MOX fuel fabrication can vary manifold. In the range between 40 and 120 tHM/a, raising the production rate with a factor of k reduces the cost of fuel fabrication at ${1 \over k}$+𝛼 rate. Here α is relatively small surcharge which can be estimated at about several percent. Every time further reduction of the fuel costs at the same absolute value requires a larger increase of the production rate which is limited by the fuel demand and the scale of nuclear power programme.
The same MOX pellet fabrication technology is normally used for thermal reactors and fast reactors. Fast reactor fuel assembly materials and manufacturing technologies differ from thermal reactors; however, this difference is expected to be relatively small. For the purpose of this assessment study it was assumed that the cost data on MOX fuel fabrication for the light water reactors will remain valid for the core fuel of fast reactors using MOX.

Table 4. Levelized cost of fast reactor closed fuel cycle services (MOX fuel, 0% discount rate)

Fuel cycle service	Cost (mills/kWh)
	BN-1200	BN-800
Core and axial blanket fuel fabrication	4.47	7.20
Radial blanket fuel fabrication	0.18	0.09
Transportation of fresh fuel	0.19	0.23
Transportation of spent fuel	0.19	0.23
Interim storage of spent fuel	0.02	0.02
Spent fuel reprocessing	1.42	1.83
Radioactive waste final disposal	1.58	2.04
Total fuel cycle cost	8.05	11.64

Table 5. Levelized cost of fast reactor closed fuel cycle services (MOX fuel, 5% discount rate)

Fuel cycle service	Cost (mills/kWh)
	BN-1200	BN-800
Core and axial blanket fuel fabrication	5.22	7.33
Radial blanket fuel fabrication	0.23	0.10
Transportation of fresh fuel	0.23	0.25
Transportation of spent fuel	0.12	0.20
Interim storage of spent fuel	0.02	0.03
Spent fuel reprocessing	0.94	1.46
Radioactive waste final disposal	0.09	0.13
Total fuel cycle cost	6.85	9.50

Due to the different content of fissile material in the spent fuel of fast and thermal reactors the safety requirements at reprocessing can be different and the associated costs of reprocessing can be different either. However, at this early stage of fast reactor deployment, with their number remaining lower than existing pool of thermal reactors and the production rate of fast reactor fuel reprocessing being lower than necessary for competitive costs, one of the effective reprocessing scenarios uses a blending of spent fuel from fast and thermal reactors. In this scenario, the concentration of fissile material in the spent fuel mix can be maintained within safety limits of the reprocessing facility nominally seen in the spent fuel from thermal reactors.
This scheme could keep the cost of fast reactor spent fuel reprocessing close to the cost of this service for thermal reactors. Previous discussion on the dependence of fuel cycle service cost on the facility production rate remains valid for the case of fuel reprocessing. More detailed consideration of the relation between the fuel cycle costs and production rates was provided in Ref. .
Evaluation of the levelized cost of fast reactor closed fuel cycle services was performed using FCCBNN code developed in the IPPE. The results of costs calculation at 0% discount rate are seen in Table 4. Fuel fabrication, spent fuel reprocessing, and radioactive waste disposal are the principal contributors to the fuel cost. The results of calculation at 5% discount rate are provided in Table 5. All costs are discounted to the moment of fuel uploading to the reactor. In this case more than three quarter of the total fuel cost is contributed by the fuel fabrication.
The fuel cost structure of thermal reactors operated in once-through fuel cycle differs from the fast reactors operated in closed fuel cycle. The thermal reactor fuel cost normally combines costs of stages different from those in Tables 4 and 5, such as the cost of natural uranium, cost of uranium refining and conversion into hexafluoride form, cost of enrichment and cost of spent fuel direct disposal. There is normally no reprocessing in once-through fuel cycle. Principal contributors to the fuel cost in thermal reactors can be different from those in Tables 4 and 5.
Theoretically all costs of fuel cycle services and the cost of natural uranium can vary depending on the market prices. Quick and short term variations are normally not accounted for in the planning stages in economic analysis. Slow and long term variations are normally estimated through the annual escalation rates which can be added to the calculation of materials or services costs.

FIG. 7. Levelized fuel costs of thermal reactor at different natural uranium cost escalation rates.

Evaluation of the effects of cost escalation for different fuel cycle services and materials is a cumbersome task which is generally out of the scope of this study. Real costs of uranium refining, conversion, and enrichment, and fuel fabrication to vary and are assumed to be constant, e.g. the thermal reactor fuel fabrication cost, 350 USD/kgHM, is constant in these calculations. However, the primary objective of the closed fuel cycle is the reduction of natural uranium consumption per energy unit produced since it is considered as a limited resource. Long term strategic planning scenarios involve a broad range of projections and assumptions on the electricity demand, and on the share of nuclear power in the energy supply mix. The availability and cost of natural uranium in long term may both depend on the global nuclear energy system characteristics and affect that system.
The effects of potential escalation of natural uranium cost throughout the NPP lifetime onto the cost of electricity produced have been evaluated in this report. The results of VVER-TOI thermal reactor fuel cost calculation at 5% discount rate are presented in Figure 7. Calculations have been performed at the different natural uranium cost escalation rates using the FCCVVR tool developed in IPPE.
Levelized costs of the fuel cycle services which do not depend on the cost of uranium and its escalation rate have been evaluated as follows:
- Front end cost:

• Refining and conversion of uranium ~ 0.3 mills/kWh;
• Enrichment of uranium ~ 2.4 mills/kWh;
• Fuel manufacturing ~ 1.0 mills/kWh;

- Back end cost (1.2 mills/kWh):

• Spent fuel reprocessing ~ 1.0 mills/kWh;
• Radioactive waste management ~ 0.1 mills/kWh;
• Transportation ~ 0.1 mills/kWh.

The effects of different escalation rates of the natural uranium cost on the levelized cost of natural uranium, levelized cost of front end and total levelized fuel cost include the following. The uranium cost annual escalation rate of 1% corresponds to the growth of cost from the accepted initial value of 100 USD/kg (in 2015) to approx. 180 USD/kg during the NPP lifetime (60 years) and the average cost of approx. 140 USD/kg. Levelized cost of natural uranium rises with a factor of 7 when escalation rate changes from 0 to 5%. At the same time the total fuel cost triples. At 0 escalation rate the cost of natural uranium contributes approx. 1/3 of the total fuel cost. At 5% escalation rate the share of natural uranium cost contribution exceeds 3/4.

Evaluation of INPRO sustainability assessment indicators in the area of economics

Calculation of basic economic functions presented in this section has been performed with the NEST tool .

Input data for cost calculation

Input data used in the INPRO sustainability assessment of BN reactors are summarized in Tables 6–8. Basic design data are presented in Table 6, investment characteristics including construction schedule – in Table 7, closed fuel cycle characteristics of BN reactors – in Table 8. Annual operation and maintenance costs of BN-1200 are equal to 122 USD/kW(e) and BN-800 – 134 USD/kW(e). Regular uniform contributions to the decommissioning fund are spread over the reactor lifetime and included in the operation and maintenance costs.

Table 6. Basic design data of BN reactors

No.	Characteristic	units	BN-1200	BN-800
1	Electric power rating	MW(e)	1220	880
2	Thermal efficiency	%/100	0.436	0.420
3	Load factor	Load factor	0.90	0.85
4	Lifetime	a	60	60

Table 7. Investment characteritics

No.	Characteristic	units	BN-1200	BN-800
1	Discount rate	%/100	0.05	0.05
2	Project overnight investments	$/kW(e)	3200	3800
3	Construction time	years	6	6
4	Investment schedule. Year of construction
	-1	%/100	0.03	0.03
	-2	%/100	0.1	0.1
	-3	%/100	0.28	0.28
	-4	%/100	0.38	0.38
	-5	%/100	0.03	0.03
	-6	%/100	0.18	0.18

Table 8. Fuel cycle characteristics

No.	Characteristic	units	BN-1200	BN-800
1	Core fuel fabrication cost	$/kgHM	3500	3500
2	Blanket fuel fabrication cost	$/kgHM	300	300
3	MOX fuel transportation cost	$/kgHM	100	100
4	Spent fuel transportation cost	$/kgHM	100	100
5	Spent fuel storage cost	$/kgHM	14	14
6	Spent fuel reprocessing cost	$/kgHM	770	770
7	Cost of radioactive waste disposal	$/kgHM	860	860
8	Core fuel burnup	MWd/kgHM	152	104
9	Time from spent fuel discharge from reactor till reprocessing (including storage in spent fuel pool)	a	5	5
10	Time from spent fuel discharge from reactor till MOX fuel fabrication	a	6	6
11	Time from spent fuel discharge from reactor till HLW final disposal (including HLW storage)	a	55	55
12	Plutonium losses at reprocessing	%/100	0.001	0.001

Cost of depleted uranium was assumed equal to zero in the calculation.

Total cost of electricity production

The results of calculation of the Levelized Unit Energy Cost are demonstrated in Figure 8. All costs are calculated for a twin unit NPP of a given type. Three major contributors to the total electricity cost are presented in different colours. Capital investment costs are blue, operation and maintenance costs are red, fuel costs are green. Costs are calculated for BN-800, BN-1200 and VVER-TOI reactors.
Costs of VVER-TOI reactor are calculated for six different assumed values of the natural uranium cost escalation rates. VVER-TOI (0) case corresponds to 0% escalation rate, VVER-TOI (1) – 1% escalation rate, VVER-TOI (2) – 2% escalation rate, VVER-TOI (3) – 3% escalation rate, VVER-TOI (4) – 4% escalation rate, VVER-TOI (5) – 5% escalation rate.
Cost of electricity from BN-800 amounts approx. 59 mills/kWh and from BN-1200 ~46 mills/kWh. Cost of electricity from the thermal reactor VVER-TOI at the constant cost of natural uranium 100 USD/kg is lower than from BN-1200. Cost of VVER-TOI energy grows depending on the escalation rate of natural uranium cost. At 2% escalation rate the cost of energy from thermal reactor may exceed the estimated cost of BN-1200.

FIG. 8. Contributors to the total electricity costs of BN-800, BN-1200 and VVER-TOI at different uranium price escalation rates.

Figures of merit

Figures of merit calculated in this sustainability assessment study include the internal rate of return, return on investment, net present value and total investment. Calculation of these parameters requires information on the price of electricity to be sold.
In the Russian Federation the price of electricity is normally determined through the bidding between electricity suppliers and customers in a given geographical and price zone. Prices include expenses necessary for electricity transfer and a surcharge for deployment of new power plants. In this study the price of baseload electricity sold by the NPP is evaluated at 52.8 mills/kWh value.

Table 9. Figure of merit

Functions	BN-1200	VVER-TOI at different natural uranium price escalation rates
		0	0.01	0.02	0.03	0.04	0.05
IRR,%/100	0.06	0.09	0.09	0.09	0.08	0.07	0.04
ROI,%/100	0.07	0.09	0.08	0.08	0.07	0.06	0.05
NPV,USD/kW(e)	1020	1530	1370	1060	750	120	-810
Total investment,106 USD	4676	4641	4641	4641	4641	4641	4641

This price of electricity, 52.8 mills/kWh, is lower than the energy cost of BN-800, 59 mills/kWh. In this case the internal rate of return of BN-800 is lower than the discount rate, net present value is negative, and the quantitative results of calculation have limited value.
Internal rate of return, return on investment, net present value and total investment for BN-1200 and VVER-TOI have been calculated with NEST tool and results are presented in Table 9. Both IRR and ROI of the BN-1200 reactor are higher than the discount rate value used for cost calculations (0.05) which indicates that this reactor theoretically can be attractive for the investor. At the uranium cost escalation rates higher than approx. 2% the NPV of BN-1200 may exceed the NPV of VVER-TOI. The BN-1200 ROI may exceed the ROI of VVER-TOI at the uranium price escalation rates higher than 3%. At the uranium escalation rates higher than 4% the IRR of BN-1200 may exceed the IRR of VVER-TOI.

Sensitivity pf electricity cost to the input data variation

Input data available at the early stages of development/ deployment of a nuclear energy system normally involve significant uncertainty. Some of the potential contributors to that uncertainty, e.g. effects of the economy of scale in different facilities, have been discussed above. Sensitivity of the levelized unit energy cost to the selected input parameters was studied using the cost sensitivity diagrams approach developed in INPRO.
Within this approach the sensitivity is defined as the levelized unit energy cost response to the variations of a selected parameter. Both the variations of selected parameter and the energy cost response are calculated in relative values which allows to compare effects and to plot several functions in one diagram. A given input parameter is varied while the rest of inputs are maintained constant. The cost sensitivity diagrams are calculated in four steps:

• Calculation of LUEC for a set of input data to be studied;
• Selection of input parameters to be varied (normally these parameters are selected among the most sensitive in different input categories);
• Calculation of LUECs for every variation of an input parameter;
• Conversion of all obtained LUECs into relative values by dividing by the value of LUEC calculated for original set of input data (see first bullet).

In this report the original value of LUEC (~46 mills/kWh) for the BN-1200 reactor has been calculated. The following parameters have been selected for the sensitivity study:

• Discount rate;
• Overnight cost;
• Operation and maintenance cost;
• Core fuel fabrication cost.

This set of parameters determines the BN-1200 electricity cost value calculated above. Discount rate is normally the most sensitive parameter of the NPP electricity cost. It affects many cost constituents including fuel costs and costs of individual fuel cycle services, however the most significant impact it provides is on the levelized unit lifecycle amortization cost which defines the interest to be paid on the investments. The value of discount rate depends on many conditions including the macroeconomic characteristics of a given country, national energy policy, the type of investor, energy market rules, risks associated with investment etc. Discount rates can be different for the same type of reactor planned to be constructed in different countries. They can be different for different types of power plants planned to be constructed in the same country. Some countries may provide a guidance on the evaluation of discount rates for different investment projects and in other countries the evaluation of discount rate can be based on academic studies.

FIG. 9. Sensitivity of the electricity cost to the input parameters.

In this assessment study, the operation and maintenance costs were presented as a single parameter. Sensitivity of the energy cost was evaluated using this aggregated parameter.
In this assessment study, the MOX fuel manufacturing is the largest single service contributor to the fuel cost of BN-1200. Theoretically, changes in fuel cycle services cost data can recast the shares of their contribution to the fuel cost, however the shape of sensitivity curve is not expected to alter significantly.
The results of the sensitivity study are presented in Figure 9.
The growth of discount rate from 5% to 20% increases the levelized cost of electricity by a factor of three. Overnight cost is the second most sensitive parameter and its quadrupling raises the cost of energy by a factor of approx. 2.6. Operation and maintenance costs are less sensitive than discount rate and overnight cost. It should be noted that the MOX fuel fabrication cost is the least sensitive parameter in this group of four.

Basic results of BN-1200 sustainability assessment in the area of economics

INPRO Economic Basic Principle: Energy and related products and services from nuclear energy systems shall be affordable and available.

User requirement UR1: Cost of energy

The cost of energy supplied by nuclear energy systems, taking all relevant costs and credits into account, C_N, should be competitive with that of alternative energy sources, C_A, that are available for a given application in the same time frame and geographic region/jurisdiction.

Criterion CR1.1: Cost competitiveness.

Indicator IN1.1: Cost of energy.
Acceptance limit AL1.1: $C_{N}\leq k\cdot C_{A}$ (C_N = cost of nuclear energy, and C_N = cost of energy from alternative source; factor k is usually ≥1 and is based on strategic considerations). The cost of electricity from BN-800, BN-1200 and advanced water cooled reactor at different costs of natural uranium is discussed in section 3.3.2.2. BN-800 generates energy at the highest cost41 among the considered reactors. The water cooled reactor has the lowest cost of electricity at the cost of natural uranium of 100 USD/kg. However, the difference between BN-1200 and water cooled reactor is very low (approx. 5% of the cost). At the uranium cost escalation rate higher than 2% the most affordable electricity may be produced by BN-1200.

User requirement UR2: Ability to finance

The total investment required to design, construct, and commission nuclear energy systems, including interest during construction, should be such that the necessary investment funds can be raised.

Criterion CR2.1: Attractiveness of investment.

Indicator IN2.1: Financial figures of merit.
Acceptance limit AL1.2: Figures of merit for investing in a nuclear energy system are comparable with or better than those for competing energy technologies.
The internal rate of return, return on investment and net present value of BN-1200 and advanced water cooled reactor at different costs of natural uranium are discussed in section 3.3.2.3. The water cooled reactor has the highest figures of merit at the cost of natural uranium of 100 USD/kg. However, the internal rate of return and return on investment of BN-1200 are higher than the discount rate which means that BN-1200 theoretically can generate reasonable profit. The BN-1200 return on investment may exceed water cooled reactor at the uranium cost escalation rates higher than 3%. The BN-1200 internal rate of return may exceed water cooled reactor at the uranium cost escalation rates higher than 4%. The BN-1200 net present value may exceed water cooled reactor at the uranium cost escalation rates higher than approx. 2%. Figures of merit for investing in BN-1200 and closed fuel cycle technology are comparable with those for water cooled reactors at the uranium cost escalation rates higher than 2%.

Criterion CR2.2: Attractiveness of investment.

Investment limit. Indicator IN2.2: Total investment.
Acceptance limit AL2.2: The total investment required should be compatible with the ability to raise capital in a given market climate.
The total investment of BN-1200 and advanced water cooled reactor are discussed in section 3.3.2.3. The difference between total investment in BN-1200 and water cooled rector is lower than 1%.
There are 36 commercial reactors operating in the Russian Federation. Several more commercial reactors are being constructed or planned. Rosatom plans to construct more than 30 commercial reactors abroad and in a few cases to invest its own funds into those international projects. The total investment required is compatible with the ability to raise capital in the Russian Federation market climate.

User requirement UR3: Investment risk.

The risk of investment in nuclear energy systems should be acceptable to investors.

Criterion CR3.1: Maturity of design.

Acceptance limit AL3.1: Technical development and status of licensing of a design to be installed or developed are sufficiently mature.
At the moment of BN-1200 sustainability assessment the decision on reactor construction has not been made and the information on status of licensing was not available to the assessor. Assessment against this criterion is not complete.

Criterion CR3.2: Construction schedule.

Indicator IN3.2: Times for construction and commissioning used in economic evaluation are sufficiently accurate, i.e. realistic and not optimistic.
In this assessment study the construction period of BN-1200 reactor was assumed to be 6 years. The prototype and demonstration reactors (BN-600 and BN-800) construction periods lasted longer than ten years in every case, however both reactors were used for the technology development including development of construction technologies. Improvements of the BN-1200 construction technologies are discussed in section 4.3.1.3 and more information can be found in the referenced materials. Ref. states that BN-1200 reactor concept enables construction of ‘N-th-of-a-kind’ reactor in 48 months. In the case of BN-1200, shorter construction periods than for BN-600 and BN-800 can be reasonably expected, however the accuracy of construction time estimation cannot be assessed in this study.

Criterion CR3.3: Uncertainty of economic input parameters.

Indicator IN3.3: Sensitivity to changes in selected parameters is acceptable to investor.
The sensitivity analysis of important input parameters for calculating costs and financial figures of merit has been discussed in section 3.3.2.4. The most sensitive parameter is the discount rate which mostly depend on the investment conditions and interest rate associated with invested funds. The investment conditions in the State corporation Rosatom differ from the free market and private investors conditions. In the case of BN-1200 construction the discount rate changes can be expected to be moderate if any.
The basic cost structure of the BN-1200 energy is similar to the basic cost structure of water cooled reactor (Figure 8) and the sensitivity of the BN-1200 overnight capital cost and operation and maintenance cost is very close to those of water cooled reactor. Sensitivity of the fuel cycle services costs is relatively low and the effect of potential increase of these costs can be expected to be reasonably low. Sensitivity to changes in selected parameters is deemed to be acceptable to investor.

Criterion CR3.4: Political environment.

Indicator IN3.4: Long term commitment to nuclear option. Acceptance limit AL3.4: Commitment sufficient to enable a return on investment.
Long term commitment of the Russian Federation government to nuclear option and support of the sodium cooled fast reactor programme was discussed in section 3.2. Commitment is deemed sufficient to enable a return on investment.

User requirement UR4: Flexibility.

Innovative nuclear energy systems should be compatible with meeting the requirements of different markets

Criterion CR4.1: Flexibility.

Indicator IN4.1: Are the innovative nuclear energy system components adaptable to different markets?
Acceptance limit AL4.1: Yes.
The BN-1200 is designed to use MOX or nitride fuels in a closed nuclear fuel cycle. In the case of domestic operation of BN-1200 or when nuclear fuel cycle facilities such as reprocessing and fabrication of nuclear fuel that use reprocessed fissile materials are located in the same country, the reactor can be fueled by different fuels providing additional benefits and flexibility to the operator. However, scenarios with import and export of closed fuel cycle services may require additional consideration of the potential proliferation resistance issues which are out of the scope of this assessment study.
The power rating of BN-1200 was selected to fit the Russian Federation electricity grid. Compatibility of the BN-1200 with small grids will need study on an ad hoc basis when necessary. Information on the BN-1200 design characteristics for reactor operation in non-baseload regime has not been available to the assessor. Assessment against this criterion is not complete.

INPRO sustainability assessment in the area of reactor safety

Introduction

INPRO basic principle for sustainability assessment in the area of nuclear reactor safety: The safety of the planned nuclear installation is superior to that of the reference nuclear installation such that the frequencies and consequences of the accidents are greatly reduced. In the event of an accident, off-site releases of radionuclides are prevented or mitigated so that there will be no need for public evacuation. The INPRO methodology has developed seven INPRO user requirements for nuclear energy system sustainability assessment in the area of reactor safety to specify in more detail the main measures presented above. The role of the INPRO assessor is to check, based on evidence provided by the designer, whether the designer has implemented the necessary measures as required by the INPRO methodology. The assessor’s product is therefore not an assessment of compliance with the IAEA Safety Standards but rather a sustainability assessment against the INPRO user requirements and criteria.

Overview of the application of the INPRO methodology area of safety to the fast reactors under development

INPRO methodology states that assessment can be carried out by a technology developer at any stage of the development of an advanced reactor design. Limited scope INPRO assessments can be focused on the specific areas and specific installations in a nuclear energy system. A developer can check whether its design under development meets the INPRO methodology sustainability criteria regarding nuclear safety. Limited scope studies may assess reactor designs under development and may help to highlight gaps to be closed by on-going R&D studies and to define the scope of data needed for making a future judgement on system sustainability. Design modifications can be initiated during early stages of development if necessary, to improve the safety level of its design. The extent and available level of detail of design and safety assessment information will increase as the design of an advanced reactor progresses to the detailed design.
The amount of information available will be a significant factor in the uncertainty of the long term validity of conclusions drawn on whether an INPRO methodology criterion has been met by the advanced design. INPRO methodology in the area of reactor safety sets up the goal to be achieved by the sustainable nuclear energy system as ‘superiority of the reactor safety to the safety of reference reactor and prevention / mitigation of off-site releases so that there will be no need for public evacuation. The user requirements specify how to achieve the goal:

• The nuclear reactor assessed is more robust than a reference design with regard to operation and systems, structures and components failures (robustness of design during normal operation);
• The nuclear reactor assessed has improved capabilities to detect and intercept deviations from normal operational states in order to prevent AOOs from escalating to accident conditions (detection and interception of AOOs);
• The frequency of occurrence of DBAs in the nuclear reactor assessed is reduced. If an accident occurs, engineered safety features are able to restore the reactor to a controlled state, and subsequently to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention is minimal, and only required after a sufficient grace period;
• The frequency of an accidental release of radioactivity into the containment / confinement is reduced. If such a release occurs, the consequences are mitigated, preventing or reducing the frequency of occurrence of accidental release into the environment. The source term of the accidental release into the environment remains well within the envelope of the reference reactor source term and is so low that calculated consequences would not require evacuation of the public; *An assessment is performed to demonstrate that the defence in depth (DID) levels are more independent from each other than in the reference design. To excel in safety and reliability, the nuclear reactor assessed strives for better elimination or minimization of hazards relative to the reference design by incorporating into its design an increased emphasis on inherently safe characteristics and/or passive systems, when appropriate;
• Safe operation of the nuclear reactor assessed is supported by accounting for human factor requirements in the design and operation of the plant, and by establishing and maintaining a strong safety culture in all organizations involved;
• The development of innovative design features of the nuclear reactor assessed includes associated research, development and demonstration (RD&D) to bring the knowledge of plant characteristics and the capability of analytical methods used for design and safety assessment to at least the same confidence level as for operating plants.

The developer of sustainable energy technology is expected to meet these user requirements. INPRO assessor employs 28 criteria which can be used for checking the status of nuclear energy system in relation to the INPRO user requirements. Four of those criteria are divided in 21 evaluation parameters for clarity. The assessor procures a total of 45 items for the INPRO assessment in the area of reactor safety:
- Design of normal operation systems:

• Margins of design;
• Design simplification;
• Improved fabrication and construction;
• Improvement of materials;
• Redundancy of operational systems;

- Reactor performance:

• Margins of operation;
• Reliability of control systems;
• Ageing management;
• Impact from incorrect human intervention;
• Sufficient technical documentation;
• Appropriate training programmes;
• Plant management organization;
• Use of worldwide operating experience;

- Inspection, testing and maintenance;
- Failures and deviations from normal operation;
- Occupational dose;
- Instrumentation and control (I&C) system and inherent characteristics:

• Continuous monitoring of plant health;
• Capability of I&C system;
• Compensation of deviations from normal operation;

- Grace periods after AOOs;
- Inertia;  Frequency of DBAs;
- Grace period for DBAs;
- Engineered safety features;
- Barriers;
- Subcriticality margins;
- Frequency of release into containment / confinement;
- Robustness of containment / confinement design;
- Accident management;
- Frequency of accidental release into environment;
- Source term of accidental release into environment;
- Independence of DID levels;
- Minimization of hazards:

• Stored energy;
• Flammability;
• Excess reactivity in the core;
• Reactivity feedbacks;
• Criticality outside the reactor core;

- Passive safety systems;
- Human factors;
- Attitude to safety;
- Safety basis and safety issues;
- RD&D;
- Computer codes;
- Novelty;
- Safety assessment.
Several INPRO criteria in this area require comparing the reactor under assessment to the reference design. In the INPRO methodology area of reactor safety, a reference reactor (or design) is a reactor of the latest design operating in 2013. It should preferably be designed by the same corporate designer as the reactor assessed and using the same technology. Based on previous experience with INPRO assessments, the definition of date for the selection of the reference design helps to avoid potential misinterpretations of terms. Note that 2013 was the date selected at the beginning of the latest methodology update. This date should be revised periodically along with the rest of the INPRO methodology.
The reference design selected for this assessment, the BN-800 reactor in Beloyarsk NPP, was commissioned in 2016. This may create a few challenges for the INPRO study of the BN-1200 reactor. For example, the accumulated operating experience of BN-800 may be insufficient for the assessment of reactor performance. In such cases the data from BN-600 can be used to support the reference reactor data.
At the time of this report preparation the BN-1200 design has not been fully optimised and finalised. Further improvement of the safety characteristics and more detailed data on the reactor characteristics could be reasonably expected. However, input data compiled in the framework of this study and presented in this report are deemed to be sufficient for the limited scope INPRO assessment to inform the developers of BN-1200 on the actions to be taken and criteria to be met to achieve the system sustainability. The assessment (self-assessment) is expected to be completed in the future along with the development of detailed design of the reactor.

UR1: Robustness of design during normal operation

INPRO methodology user requirement UR1 for sustainability assessment in the area of safety of nuclear reactor was formulated as follows: “The nuclear reactor assessed is more robust than a reference design with regard to operation and systems, structures and components failures”.

Criterion CR1.1: Design of normal operation systems

Indicator IN1.1: Robustness of design of normal operation systems.
Acceptance limit AL1.1: More robust than that in the reference design.

Evaluation parameter EP1.1.1: Margins of design

The BN-1200 and BN-800 are large pool type sodium cooled fast reactors. Both in the BN 1200 and BN-800 the reactor core, primary radiation shielding, primary coolant pumps and intermediate heat exchangers are located within the main reactor vessel. Main reactor vessel is enclosed in the guard vessel. Basic characteristics of BN-1200 and BN-800 reactors are provided in Table 10.
BN-800 reactor uses hexagonal fuel assemblies with 96 mm flat-to-flat outside duct size (fuel assembly pitch - 100 mm). The diameter of fuel pins in all assemblies is the same and amounts 6.9 mm. BN-1200 uses enlarged hexagonal fuel assemblies with 181 mm flat-to-flat outside duct size and fuel pins of different diameters - 9.3 mm and 10.5 mm . Different fuel pins are used for designing required power distribution in the reactor core. Besides that, fuel modifications in the BN-1200 reactor involve enhancement of the gaseous plenum for accumulation of fission products at higher burnups.
Due to these modifications the volumetric power density in the BN-1200 reactor core has been reduced by a factor of 1.8 compared against BN-800. Ref. estimates average power density in BN-800 reactor core at 400-450 MW/m3 and in BN-1200 – at 237-255 MW/m3 . Maximum linear heat generation rate in BN-800 reactor core fuel is 48.5 kW/m and in BN-1200 – 46.5 kW/m.
Melting point of the mixed oxide (MOX) fuel is 2750°C. At normal operation of the BN-800 reactor the fuel temperature may reach 2500°C with a corresponding safety margin of 250°C. The highest fuel temperature that the BN-1200 reaches is 2100°C and the fuel temperature safety margin of 650°C.

Table 10. Basic characteristics of BN-800 and BN-120 REACTORS

Characteristics	Reactors
	BN-800	BN-1200
Thermal power rating, MW(th)	2100	2800
Design lifetime of non-replaceable equipment, a	40	60
Specific metal content of the reactor, t/MW(th)	4.1	2.4
Number of coolant circuits 3 3 Primary coolant temperature, in/out, °C	354/547	410/550
Secondary coolant temperature, in/out, °C	309/505	355/527
Tertiary coolant temperature, in/out, °C	210/490	275/510
Number of loops in every circuit	3	4
Steam generator design	modular	shell and tube type
SG outlet steam pressure, MPa	13.7	17
Steam reheating steam steam Electric power rating, MW(e)	880	1220
NPP efficiency, gross/net, %	41.9/38.8	43.6 /40.9
Turbine type	K-800-130	K-1200-130
Generator type	TZV-800-2	TZV-1200-2
Estimated load factor, %	85	90

Maximum fuel cladding temperature at normal operation conditions in all BN family reactors is limited by 710°C. In different zones of the BN-800 reactor core the fuel claddings may reach 670 - 690°C providing a safety margin of 20 °C. In BN-1200 at normal operation the claddings temperature does not exceed 680°C in any part of the reactor core and the safety margin amounts at least 30°C.

Evaluation parameter EP1.1.2: Design simplification

Unlike other members of the BN family of reactors, the BN-1200 design layout is symmetric and the layout of all four loops in every circuit is identical which simplifies the manufacturing and construction of reactor systems structures and components.
BN-800 secondary circuit scheme including steam generators and EHRS is presented in Figure 10. Three loops are connected to the reactor vessel and 60 modules of steam generators. BN-1200 technology scheme is presented in Figure 11. Four identical loops (located in identical premises at the same levels) and EHRS modules are connected to the reactor vessel in the centre.
Using bellows type compensators in secondary circuit of BN-1200 instead of compensating elbows as it was done in BN-800 resulted in reducing of the total length of secondary sodium pipelines from ~770 m to ~400 m. The BN-1200 design reduced the number of fittings by about 930. Table 11 provides a comparison of fittings used in the BN-800 and BN-1200 reactors.
BN-1200 design includes 8 vertical steam generators of ‘shell and tube type’ instead of 60 modules of ‘modular type’ steam generators in BN-800. Unlike BN-800 the level of sodium in BN-1200 reactor does not reach the conical part of the reactor vessel which eliminates the necessity of guard vessel and bellows in this part of reactor.

FIG. 10. BN-800 secondary circuit scheme

FIG. 11. BN-1200 technology scheme

Table 11. Fittings used in BN-800 and BN-1200.

Fittings	BN-800	BN-1200
	Sodium fittings
DN10-DN100	870	370
DN-300	75	-
DN-400	-	20
	Steam/ water fittings
DN10-DN100	438	130
DN175-DN300	102	35

BN-800 reactor core contains 565 fuel assemblies having three different initial fuel composition (different content of fissile material). The 432 fuel assemblies of BN-1200 reactor core have bigger cross-sections and initially have uniform fuel composition which simplifies the reloading procedures.
Increasing the capacity of in-vessel short term spent fuel storage allowed storing the spent fuel for 2 years and to reduce residual heat density from a spent fuel assembly to 2 W/cm³. These low residual heat densities avoid needing external ‘drum’ storage (using sodium for cooling of spent fuel assemblies) and a number of related auxiliary systems (for cooling, draining, sodium cleaning, protection from high pressure, oxides monitoring, cooling of oxides monitoring probes, etc). The BN-1200 fuel reloading system design includes the vertical elevator, combined reloading/washing section and other features differing from the BN-800. Using these features the BN-1200 reactor fuel reloading system was further simplified and the mass of steels and other materials (per kW of reactor power rating) used for the system manufacturing was reduced by a factor of 10.
The BN-800 reactor was designed for using MOX fuel. The BN-1200 reactor design plans for use of MOX fuel only in early stages of the programme. It is expected that at the more advanced stages when nitride fuel technology is validated BN-1200 reactors will operate with nitride fuel. The 174 blanket assemblies around the BN-1200 reactor core fuelled with nitride fuel will be replaced with steel/boron-carbide shielding that strongly reduces the neutron irradiation of in-vessel structures and components. This allows elimination of in-vessel shielding which is used in the BN-800 reactor.

Evaluation parameter EP1.1.3: Improved fabrication and construction

BN-1200 reactor building is designed against new and more strict regulatory requirements. External structures of the building are reinforced, and the expansion joint has been introduced along the confinement walls to prevent the transmission of harmful impact to the reactor systems structures and components in case of an accident. A new relatively light spherical dome is envisioned to be placed above the reactor. The dome is to be assembled using the modules fabricated from the modern high-strength reinforced concrete which increases the efficiency of construction process.
Other expected improvements in the construction of BN-1200 are mostly related to the reactor layout. All four loops of secondary cooling circuit are identical and corresponding systems structures and components are located symmetrically at the same levels which allows to use an approach conceived for construction of a modular facility.
The Russian architect engineering company NIIAEP has developed an original ‘Multi-D’ digital technology which can be used at the design, construction and operation stages of the reactor lifecycle . It creates a complete universal 3D (design + construction + technology) model of the object, controls potential discrepancies and collisions, manages the configuration at any details level, communicates among users and prepares working level documentation. This information system allows modelling of all basic construction processes, preparation of detailed graphs and involves several modes of in-field application. If this tool confirms its effectiveness it can theoretically help to reduce a potential negative impact of human factor during the construction of BN-1200.
The BN-1200 major equipment fabrication methods are not expected to differ from those used for BN-800. The reactor vessel assembly will be completed on site. The only expected difference is the necessity of on-site assembling of the large rotating plug of BN-1200, however this procedure cannot be considered as an improvement of fabrication process.

Evaluation parameter EP1.1.4: Improvement of materials

The BN-1200 design involves a number of modifications of structural material in fuel pins, reactor internals, steam generators, anti-cavitation coatings, etc. The designer used new high-strength stainless steel 07Cr12NiMoVNb instead of 10Cr2Mo in BN-800. to increase the durability and lifetime of the BN-1200 steam generators. The lifetime of the BN-800 steam generator module reaches 150 000 hours. BN-1200 steam generator’s lifetime extends to 240 000 hours .
Several selected in-reactor systems, structures and components of BN-1200 will be fabricated using the new heat and radiation resistant steel Cr16Ni11Mo3. This material allows an increase of system’s lifetime up to 60 years and improve safety characteristics of the reactor.
BN-600 and BN-800 reactors use austenitic steel EK-164 for fuel claddings. BN-1200 uses both austenitic steel EK-164 and ferritic-martensitic steels EK-181 and ChS-139 as fuel cladding materials which improves safety features (higher heat resistance, higher damaging dose) economics of fuel management (higher burnups).
As noted above, the BN-800 reactor can operate with both UOX and MOX fuel. The BN-1200 reactor will use MOX fuel only in early steps of the programme with nitride fuel expected for later use. Nitride fuel provides higher density of fuel material and improves safety and economics characteristics of reactor
New effective steel/boron-carbide shielding surrounding the core strongly reduces the neutron irradiation of reactor vessel and in-vessel structures and components. This allowed a reduction of the steel shielding located within the BN-1200 reactor from 830 t to 27 t.

Evaluation parameter EP1.1.5: Redundancy of operational systems

It is acknowledged that an increase of redundancy may increase the complexity of a system as discussed in evaluation parameter EP1.1.2 above. Thus, the design needs optimization with respect to system redundancy and complexity. Design simplification generally cannot be used as justification for reducing the redundancy of operational systems.
Data demonstrating that the redundancy of operational systems is greater than that in the reference design have not been available to the INPRO assessor. Assessment against this evaluation parameter is not complete.

Criterion CR1.2: Reactor performance

Indicator IN1.2: Reactor performance attributes.
Acceptance limit AL1.2: Superior to those of the reference design.

Evaluation parameter EP1.2.1: Margins of operation

Prior to the assessment of this evaluation parameter, it is important to note that prototype reactor BN-600 has relatively high performance characteristics compared to the currently operated water cooled reactors. Ref. presents evaluation of scram frequency in BN-600 (0.2 per 7000 hour of operation) compared to global average value (0.5–0.7 per 7000 hours per power unit). There were no scrams in BN-600 in the period from 2000 through 2013.
Improvement of operational margins in BN-1200 reactor is mostly related to the reduction of the maximum power density of the core to 380 MW/m³ which is lower than average power density in BN-800 and BN-600 reactors.
Ref. provides the list of 12 normal operation states in BN-1200:

• 5 working states including reactor operation at different power levels, reactor start-up and shutdown from/to different states (hot or cold);
• 5 shutdown states including fuel reloading, steam generators chemical cleaning, etc;
• working in a mode providing energy for own needs only;
• manoeuvrable mode of operation.

Ref. presents a list of normal operation levels and margins identified for BN-1200. Normal operation levels for primary coolant temperature at the outlet from the core and secondary coolant temperature at the reactor outlet can be compared with the nominal values of these parameters to estimate corresponding margins:

• Primary coolant temperature at the outlet from the core 550-600°С;
• Secondary coolant temperature at the reactor outlet 527-550°C;
• Pressure in the reactor gas cavity – 0.005-0.054 MPa;
• Gas pressure in the secondary circuit – 0.25-0.28 MPa.

Other normal operation levels provided in Ref. [50] are as follows:

• Reactor vessel temperature (60 years) ≤ 504°С;
• Reactor components temperature change rate ≤ 30°C/hour;
• Gas activity in the reactor gas cavity ≤ 500 MBq/l;
• Primary sodium activity ≤ 50 MBq/kg;
• Fuel pins damage of gas leak type - 0.05% of all fuel pins;
• Fuel pins damage of fuel-coolant contact type – 0.005% of all fuel pins.

The normal operation levels and margins of the BN-800 reactor are not available in the open domain to allow the assessor to finalize assessment.

Evaluation parameter EP1.2.2: Reliability of control systems

Reference design, BN-800, is one of the newest and most advanced currently operated reactor designs. Development of I&C system for BN-800 was based on the contemporary requirements. Distributed control principle has been implemented in the system providing coordinated and independent processing and control functions to different system modules. This system fulfils the detailed diagnostics of reactor systems and self-diagnostics.
BN-800 reactor control and protection system is one of the I&C subsystems that combines control functions (e.g. reactor start-up, power control, shutdown etc), protective functions (emergency power level reduction, scram) and monitoring functions (core and neutron guide monitoring; systems structures and components monitoring; information exchange with other I&C subsystems). The system was designed to maintain the following sequence of control priorities:
1. Reactor protection (scram);
2. Power level reduction;
3. Prevention withdrawal of control rods;
4. Manual control of reactor;
5. Automatic control of reactor.
Reactor control and protection system consists of several subsystems:

• Digital automatic reactor shutdown systems (scram);
• Analogue automatic reactor shutdown systems (scram);
• Digital control and monitoring subsystem;
• Driving gears control subsystem;
• Main control room safety panels;
• Emergency control room safety panels.

The control and monitoring subsystem design has a triple reservation and performs automated reactor control in start-up, manoeuvring and scheduled shutdown modes, carries out automatic reactor power control, emergency power reduction control, compensation of fuel burn-up effects, control of electric actuators of driving gears prior to and after reloading, monitoring and indication of the control rods position, of system structures and components status, automatic self-diagnostics of the system hardware and software in all states of the reactor.
The BN-1200 control systems have a basis in the BN-800 developments and will accommodate a number of improvements related to the reactor optimization and simplification. For example, BN-800 neutron flux monitoring system contains a neutronic guide, two sets of in-reactor ionization chambers (2×4 KNT 54-2) and 24 near-reactor ionization chambers KNK 15-1. Use of four sets of in-vessel ionization chambers for the neutron flux monitoring and elimination of neutron guides improves the reliability of control systems in BN-1200.

Evaluation parameter EP1.2.3: Ageing management

In the operating prototype fast reactor BN-600 an ageing management programme was introduced more than two decades ago. It is expected that in the BN-1200 reactor the ageing management programme will cover all steps of the project development. At the design stage, the BN-1200 designer determined the design life of items important to safety () and expected to provide appropriate design margins to take due account of age-related degradation and to provide methods and tools for assessing ageing during the NPP operation.
The detailed design of BN-1200 involves development of complex systems for diagnostics of damage and evaluation of residual lifetime of the reactor components and core structural elements.

Evaluation parameter EP1.2.4: Impact from incorrect human intervention

Data demonstrating that incorrect human intervention during normal operation has less impact on reactor operation than in the reference design have not been available to the INPRO assessment. Assessment against this evaluation parameter is not complete.

Evaluation parameter EP1.2.5: Sufficient technical documentation

The first BN-1200 reactor is planned to be constructed at Beloyarsk NPP site. There are four reactors currently located at this site including BN-600 and BN-800. Beloyarsk NPP is the oldest currently operating NPP in the Russian Federation and well-established enterprise having a very well developed documentation system. The general quality assurance programme of the Beloyarsk NPP POCAS (O), put into effect by order of national utility, Rosenergoatom JSC, in 2014. Detailed documented procedures for a quality management system have been developed and implemented at Beloyarsk NPP (titles of documents):

• Records Management at Beloyarsk NPP;
• Management of Technical Documentation for the Industrial Activities of Beloyarsk NPP; *Paperwork Manual at Beloyarsk NPP;
• Planning and Monitoring of Occupational Activities at Beloyarsk NPP;
• Maintenance and Repair at Beloyarsk NPP;
• Metrological Support at Beloyarsk NPP; *Procurement Procedures at Beloyarsk NPP;
• Human Resource Management at Beloyarsk NPP;
• Non-compliance. Corrective and Preventive Actions;
• Regulation on the Organization and Conduct of Internal Quality Audits at Beloyarsk NPP.

The quality management system of Beloyarsk NPP is subject to certification peer-reviews organized on a regular basis. In 2014, the ANO Atomcertifica (Russian independent certification body) conducted the peer-review which confirmed compliance of the quality management system of the Beloyarsk NPP to the national and international requirements.
Ref. gives a brief overview of technical documentation related to the nuclear hazardous works including relevant testing. These documents include:

• Beloyarsk Nuclear Power Plant. Final Safety Assessment Report;
• Nuclear Safety Manual for Storage, Transportation and Reloading of Nuclear Fuel of AMB-100 and AMB-200 reactors (Units 1 and 2);
• Operating Manual of First Line of Beloyarsk NPP (Units 1 and 2);
• Operating Manual of Unit 3 of Beloyarsk NPP (BN-600);
• Operating Manual of Unit 4 of Beloyarsk NPP (BN-800);
• Nuclear Safety Manual for Storage, Transportation and Reloading of Nuclear Fuel of BN600 reactor (Unit 3);
• Nuclear Safety Manual for Storage, Transportation and Reloading of Nuclear Fuel of BN800 reactor (Unit 4).

BN-1200 reactor is reasonably expected to be provided with full set of necessary technical documentation.

Evaluation parameter EP1.2.6: Appropriate training programmes

The first BN-1200 reactor is planned to be constructed at Beloyarsk NPP site where BN-600 and BN-800 reactors are accommodated. Operating experience of these two reactors will be used in the training programmes to be developed for BN-1200. Well qualified personnel from BN-600 and BN-800 operating staff can be used for training of new staff and for commissioning and operation of BN-1200.
Training facilities of BN-800 reactor include a full-scale simulator created in 2015-2019 and some of these facilities can be used for the BN-1200 personnel training.

Evaluation parameter EP1.2.7: Plant management organization

BN-800 reactor of Beloyarsk NPP was commissioned in 2016 (first criticality obtained in 2015). Russian companies that participated in the construction and commissioning of BN-800 including the designer organization (OKBM) and companies affiliated to the national utility (Rosenergoatom) have the necessary expertise for start-up and operation, and the competence of utility and Beloyarsk NPP is appropriate and meets national licensing requirements.
These organizations proved to have enough personnel, well defined structures, functions, lines of management and detailed qualification requirements/job descriptions. Beloyarsk NPP, like other Russian NPPs, has identified managers responsible for operation, maintenance, technical support, quality assurance, environmental protection, nuclear and industrial safety, and administration.

Evaluation parameter EP1.2.8: Use of worldwide operating experience

BN-1200 design is based on the experience obtained from the operation of BN-350, BN-600 and BN-800 reactors BN-1200 designer participates in the information exchange programmes related to safety of LMFR and organized by the IAEA and other international organizations. JSC OKBM has developed several types of reactors different from BN-1200 (including several types of propulsion reactors for commercial fleet and navy) and design of fuel assemblies for commercial VVER-1000 (Russian type of large PWR). However, the assessor had no direct information available on the accounting of operating experiences from foreign NPPs operating types of reactors other than fast reactors. Some experience from water cooled reactors may be applicable to the fast reactors either (e.g. the Forsmark NPP 2006 electrical event). Assessment against this evaluation parameter is not complete.

Criterion CR1.3: Inspection, testing and maintenance

Indicator IN1.3: Capabilities to inspect, test and maintain.
Acceptance limit AL1.3: Superior to those in the reference design.
The experience of operating the BN-600 power unit shows that leaks from sodium pipelines and in steam generators took place mainly in the early period of reactor operation. Leaks were mostly caused by deviations in the quality of equipment manufacturing. Pure sodium provides very low corrosive-erosive effects on the materials of systems, structures and components.
Total number of sodium leaks in BN-600 added up to 39 including 12 leaks in steam generators. The rest of the 27 leaks occurred in the auxiliary systems including 5 leaks from primary circuit systems. Six leaks exceeded 10 kg of sodium (largest mass of sodium leak obtained 1000 kg – level 1 of INES scale). One leak from the primary circuit caused a minor radioactive release to the environment (lower than regulatory limits for normal operation). Sodium fires occurred in 14 cases. The latest leakage of sodium in the system occurred in 1994 and in a steam generator in 1991 .
Ref. provides the detailed guidance on the inspection of systems, structures and components of BN-800 reactor. It determines the equipment to be controlled, control methods, scope of control and frequency. Because of the robust design features of the reactor (integral layout of the primary equipment of the primary circuit), primary and secondary coolant features (liquid sodium), double wall reactor vessel (main vessel and guard vessel) and the guard casings of primary auxiliary sodium systems pipelines, operational inspections of the base metal and welded joints are not provided for the in-vessel structures, parts of the reactor vessel and primary pipelines covered by guard vessel and guard casings.
For the rest of BN-800 systems, structures and components the inspections involve visual, capillary, ultrasonic, radiographic and magnetic particle methods.
Scope of inspections and their frequency depend on the system safety classification and on the knowledge of damage mechanisms, design specifics and operating conditions. Where the scope of a single inspection is less than 100%, different groups of welded joints should be inspected every next time. If defects are detected during the selective inspection, the scope of inspection must be doubled and the adjacent areas inspected. If further defects are detected, a 100% inspection of the system, structure or component has to occur.
Research studies on the improvement of inspection methods have been included in the R&D programme of BN-1200. Inspections in BN-1200 are expected to be more simplified and more effective due to the reactor design simplification.
Detailed information on BN-1200 design features facilitating the performance of testing and maintenance and potential improvements of their effectiveness and efficiency has not been available to the assessor. Assessment against this criterion is not complete.

Criterion CR1.4: Failures and deviations from normal operation

Indicator IN1.4: Expected frequency of failures and deviations from normal operation.
Acceptance limit AL1.4: Lower than that in the reference design.
Results of analysis of the number of failures and human errors in the prototype reactor BN-600 appears in Figure 12 (data taken from Ref.</ref>). Based on this analysis, the frequency of AOO in first-of-a-kind reactors (including BN-600, BN-800 and BN-1200) can be a time-dependent characteristic. The ratio between the AOO frequencies in the beginning of operation and 30 years later can be higher than 10.

FIG. 12. Number of failures and human errors per calendar year from the BN-600 commissioning in 1982</ref>.

Ref. states that the BN-1200 design documentation considers 64 different AOOs. The most remarkable consequences (within acceptable limits) can be caused by the following events:

• Uncontrolled withdrawal of a single control rod;
• Loss of power to a primary coolant pump;
• Loss of power to the own needs;
• Water to sodium leaks through the heat exchange tubes of steam generator.

The improvements implemented in the BN-1200 design are expected to reduce the frequency of AOOs. Most notable of them include:

• Design simplification discussed in EP1.1.2 reduced number of systems, structures and components along with pipelines length and is expected to reduce both the frequency of failures and frequency of deviations from normal operation caused by human errors;
• New layout of the entire primary circuit within the reactor vessel minimized the frequency of AOOs involving leaks of primary sodium;
• New materials improving reliability of systems structures and components will also reduce the frequency of failures.

Direct information on the frequencies of AOO in the BN-800 and BN-1200 was not available to the assessor. Assessment against this criterion is not complete.

Criterion CR1.5: Occupational dose

Indicator IN1.5: Occupational dose values during normal operation and AOOs.
Acceptance limit AL1.5: Lower than the dose constraints.
Ref. provides collective occupational dose for period of 20022015 for operating prototype reactor BN-600 – 0.408 man×Sv/a. This value is lower than average values for operating PWRs, BWRs or PHWRs provided in the INPRO manual.
Ref. provides average annual individual occupational doses for every Russian NPP between 1991 and 2003 (Table 12) and average dose for all NPPs in 2010 (1.4 mSv/a) which is lower than global average (2.4 mSv/a).

Table 12. Average annual individual occupational doses for Russian NPPs(Ms/a) .

NPP	1991	1995	1996	1997	1998	1999	2000	2001	2002	2003
Balakovo	0.8	1.4	1.0	1.0	1.2	1.0	0.8	0.7	0.7	0.7
Beloyarsk	1.8	1.6	1.6	1.3	2.2	1.4	1.8	1.7	1.6	1.0
Bilibino	9.7	8.8	11.5	6.0	6.9	5.8	4.9	5.3	5.2	4.4
Volgodonsk								0.02	0.07	0.1
Kalinin	2.3	1.8	1.5	1.4	1.2	1.2	1.2	1.0	0.7	0.6
Kolskiy	4.0	3.0	3.2	1.8	2.0	3.2	2.0	2.1	1.8	1.9
Kursk	16.0	11.0	9.8	7. 9	6.2	6.9	5.9	4.3	4.4	3.6
Leningrad	5.2	7.1	6.6	5.8	4.9	3.5	3.9	4.0	3.5	3.5
Novovoronezh	4.8	4.9	2.9	2.8	2.3	3.5	2.3	3.1	2.7	2.6
Smolensk	4.6	4.3	3.8	4.6	5.4	5.2	4.8	4.6	4.6	2.3

For the BN-1200 project, the estimated occupational doses are lower than for BN-600 and BN800 (due to the design simplification, optimization of fuel reloading procedures) and are reasonably expected to be lower than dose constraints.
The effects of potential releases of C-14 at normal operation from the BN-1200 reactor using nitride fuel are estimated in Ref.. It was found that after 10 years of operation the activity of C-14 in the primary coolant will achieve ~10¹¹ Bq and in the primary cover gas ~2×10¹⁰ Bq. Individual occupational dose from C-14 can amount ~0.02 mSv/a (compare to regulatory limit – 20 mSv/a). The normalized annual C-14 emission rate to the environment was estimated at ~0.02 TBq/GW(e)×a which is lower than typical values for water cooled reactors in Russian Federation.

UR2: Detection and interaction of anticipated operational occurences

INPRO methodology user requirement UR2 for sustainability assessment in the area of safety of nuclear reactor was formulated as follows: “The nuclear reactor assessed has improved capabilities to detect and intercept deviations from normal operational states in order to prevent AOOs from escalating to accident conditions”.

Criterion CR2.1: I&C system and inherent characteristics

Indicator IN2.1: Capabilities of the I&C system to detect and intercept and/or capabilities of the reactor’s inherent characteristics to compensate deviations from normal operational states.
Acceptance limit AL2.1: Superior to those in the reference design.

Evaluation parameter EP2.1.1: Continuous monitoring of plant health

Ref. provides the list of basic monitoring provisions in BN-600 and BN-800 reactors:

• Primary vessel and internals – under-gas viewing (inside of primary vessel of reactor) / aerosol detection of primary vessel leak and electrical contact (outer surface of primary vessel);
• Secondary circuit pipes – leak detectors (electrical contact), smoke detectors;
• Intermediate heat exchangers – control Na level and Ar pressure;
• Steam generators – leak detectors, control of Na level and Ar pressure, regular visual inspection of tube-bores and structures.

The operational manuals provide more details on the monitoring systems of the BN-800. Inreactor monitoring system performs on-line measurement and processing of the following characteristics (operation at different power levels):

• Sodium temperature in the reactor vessel, at the inlets/outlets of intermediate heat exchangers, at the outlet from fuel assemblies (37×3 probes), etc. More than a hundred detectors in total;
• Temperature and deformation of reactor vessel (in 90 locations);
• Levels of sodium in the reactor vessel and in the primary coolant pumps tanks;
• Sodium flow rate through the core;
• Gas pressures in the reactor and between the main reactor vessel and guard vessel;
• Vibrations of reactor vessel, heat exchangers and primary coolant pumps;
• Neutron flux;
• Fuel leaks (through gas fission products activity and delayed neutrons);
• Coolant boiling;
• Reactor vessel leaks.

Different characteristics are monitored during the reactor reloading. Refs provide additional information on the following monitoring systems in BN-800:

• Loose parts monitoring;
• Monitoring of rotating machinery;
• Seismic monitoring;
• Radioactive releases and environmental impact monitoring;
• Monitoring of fires (as part of fire protection system).

Ref. states that basic design principles of the BN-800 reactor I&C system are similar to those of water cooled reactors of Russian design. Peculiarities mostly occur because of using sodium as a coolant in fast reactors. Special tools for measurement and monitoring of sodium systems were developed together with sodium electric heating instrumentation and control system. In particular, the number of monitored heating areas was increased up to 3500, and temperature sensors (two thermocouples) were installed in each area in order to maintain coolant temperature within the range of 550 ± 30°C. For the purpose of sodium parameters monitoring, new sensors were developed and introduced to measure pressure, level and flow rate, monitor sodium leakage and hydrogen concentration in reference points. Changes were also introduced to I&C subsystems aimed at controlling the turbine hall parameters. In addition, I&C of Beloyarsk NPP Unit 4 performs online self-diagnostics and process equipment monitoring providing the operator with the full scope of necessary information.
These monitoring systems are expected to be incorporated in the BN-1200 design.
Direct information on the improvement of monitoring systems in BN-1200 was not available to the assessor. Assessment against this evaluation parameter is not complete.

Evaluation parameter EP2.1.2: Capability of the I&C system

Ref. gives general overview of the BN type reactors features relevant to their controllability. It is stated that BN type reactors are easier to control and more stable than existing thermal reactors and can be operated at a given power level for many hours without interventions of the control system. Basic features of BN reactors involve:

• Highly stable relative power distribution in the core with a low sensitivity to the control rods position;
• No reactor poisoning with xenon and samarium;
• Low inertia of main constituents of negative feedbacks;
• Short lags in response to the changes of basic parameters;
• Low excess reactivity.

One improvement of I&C system capability in BN-800 has been achieved through introduction of the reactor diagnostics system [68]. This unique automated system was designed for comprehensive monitoring and projection of the scenarios and processes occurring in the reactor installation under normal operation conditions and AOO at different reactor power levels (from 0.1 to 120%). Reactor diagnostics system performs analysis of the neutronic (including neutronic noise), thermal hydraulic and other technological characteristics of the reactor, early detection of deviations that could result in the substantial damage of fuel including sodium boiling, formation of ‘hot’ spots, reduction of coolant flow rate in a fuel assembly, fuel rod melting, intermediate heat exchangers coolant flow rate deviations, uncontrolled withdrawal of control rods, in-vessel vibrations, reactor core coolant flow rate deviations, I&C systems failures.
Statistical estimates of temperature distributions, reactivity values, statistical and spectral characteristics of neutronic noise temperature channels during operation at different power levels were obtained. Reactor diagnostics system uses improved methods for the measurement and improved algorithms for complex analysis of data from the abnormal reactivity detection subsystem, neutronic noise diagnostics subsystem, reactor core temperature monitoring subsystem, fuel claddings leak-tightness gas/delayed neutrons monitoring systems, and from other I&C systems giving high efficiency of AOO detection [68]. A similar system is expected to be implemented in BN-1200 reactor.
The BN-1200 design provides several other improvements of the I&C systems. Four in-reactor sets of ionization chambers for the neutron flux measurements in BN-1200 provide better quality of measurements and improve reactor control. Ref. provides a description of improved fire protection system developed for BN-1200.

Criterion CR2.2: Grace periods after AOOs

Indicator IN2.2: Grace periods until human actions are required after AOOs.
Acceptance limit AL2.2: Longer than those in the reference design. Direct information on the longer grace periods after AOO in BN-1200 design was not available to the assessor. Assessment against this criterion is not complete.

Criterion CR2.3: Inertia

Indicator IN2.3: Inertia to cope with transients. Acceptance limit AL2.3: Larger than that in the reference design. BN-1200 and BN-800 are both sodium cooled reactors. Thermal conductivity of sodium is ~300 times higher than that of liquid water. Both designs are the pool type reactors having all major47 primary circuit components and primary coolant located within the reactor vessel. Such a reactor has high heat capacity. The BN-800 reactor vessel has diameter of 12.9 m and height of 16.41 m. Volume of the primary sodium in the BN-800 amounts approx. 1100 m³. In BN-1200 these dimensions are larger – 16.9 m and 20.72 m, and the power rating / volume ratio for BN-1200 is lower than in BN-800 and corresponding characteristics of inertia can be expected to be higher.

The BN-1200 reactor possesses considerable thermal inertia. In the analysis of severe accidents scenario with station blackout, active shutdown and EHRS failures (ultimate loss of flow plus ultimate loss of heat sink type accident, ULOF+ULOHS) the process of reactor heat-up due to residual heating up to the sodium boiling temperature (900°C at a given pressure level) takes about 2 days (see Figure 13 taken with modifications from Ref.).

FIG. 13. Reactor coolant temperature in the BN-1200 accident with station blackout and EHRS failure. Line 1- sodium temperature at the bottom of reactor, Line 2 – sodium temperature at the intermediate heat exchangers inlet, Line 3 – argon temperature near the reactor lid.

Ref. states that BN-1200 design uses a new type of primary coolant pumps providing longer coast-down in the case of power loss. The transient response seen in the analysis of station blackout scenario with the failure of two trains of EHRS out of total four showed this behaviour. Loss of power results in disabling of primary and secondary coolant pumps and loss of feedwater supply to the steam generators. Reactor power drops down to the level of residual heat due scram provided by passive systems. During the primary coolant pumps coastdown (~60 s) the check valves of autonomous heat exchangers (EHRS) and outlet gates of air heat exchangers open providing natural circulation of coolants in the EHRS circuits and in reactor (for more details see below for the discussion of severe accidents). The fuel element cladding temperature reaches values close to the normal operation limit for a short period. The reactor vessel temperature remains close to the nominal value.

UR3: Design basis accidents

INPRO methodology user requirement UR3 for sustainability assessment in the area of safety of nuclear reactor was formulated as follows: “The frequency of occurrence of DBAs in the nuclear reactor assessed is reduced. If an accident occurs, engineered safety features are able to restore the reactor to a controlled state, and subsequently to a safe shutdown state, and ensure the confinement of radioactive material. Reliance on human intervention is minimal, and only required after sufficient grace period.”