Wiki-INPRO-methodology

The INPRO methodology is an IAEA tool that assists Member States in strategic planning and decision making on nuclear power programmes by sustainability assessment of a NES.

Introduction of the INPRO methodology

Application of the INPRO methodology

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:
$C_{N}\leq k\cdot C_{A}$ (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

ᅠOvernight capital costᅠ

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. 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.

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

ᅠOperation and maintenance costᅠ

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. 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.

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

{{NoteL|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.
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.
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.
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.