Core and fuel cycle for an advanced sodium-cooled fast reactor

Abstract

The development of BN-1200 is based on the greatest possible use of tested and scientifically validated and developed technical solutions implemented in , and the BN-800 design as well as new technical solutions that increase facility cost-effectiveness and safety. The BN-1200 design must permit the reactor to operate with different cores, including with denser fuel. The main fuel variant considered is oxide fuel and for the nearest term nitride fuel, for which the production technology involves the same steps as the oxide technology. The main approaches for choosing the parameters of the BN-1200 core as well as the results of computational studies are presented.An advanced sodium-cooled fast reactor satisfying the requirements for nuclear power in the 21st century is now being developed in our country. Initially, 1800 MW(e) capacity was being considered for this reactor [1]. However, for many reasons, specifically, to ensure transportability of the main equipment by rail, its capacity was lowered to 1200 MW. The development of BN-1200 is based on the maximum use of the solutions implemented in BN-350 and -600 and in the BN-800 design as well as new technical solutions which increase cost-effectiveness and safety.BN-1200 Core. The main variant considered is oxide fuel and for the nearest term nitride fuel, whose production technology is largely identical to that of oxide fuel. Variants with heterogeneous introduction of regions with depleted metallic uranium are also considered. Metallic fuel does not satisfy the BN-1200 thermodynamic indicators which have been adopted. For fuel elements to work reliably with such fuel, the maximum coolant temperature must be lowered by approximately 100°C.The high burnup adopted for the fuel (16-18% h.a.) is attained by using in the fuel-element cladding EP-450 type ferrite-martensite steel; for this, the maximum cladding temperature in the design is limited by 670°C. For such fuel-element cladding temperature, the coolant heating in the reactor is lowered (140°C). This makes it possible to lower the maximum sodium temperature at the entrance to the fuel assemblies; this temperature is associated with the nonuniformity of the coolant heating in the core and the fuel assemblies. In the future, a transition is planned to low-swelling ferrite-class high-temperature steel, dispersion hardened by oxides of rare-earth elements.An important characteristic of the core is its volume heat density, since it actually determines the core size and critical load. Initially, a high heat density (~500 MW/m 3 ) was chosen for fast reactors. This choice was made to minimize the

System levelized fuel cost of electricity generation in a two-component nuclear energy system with a closed uranium-plutonium NFC

“…Let us recall the main provisions of the traditional approach to calculating the LUFC (OECD/ NEA 1994, IAEA 2008, IAEA 2014, Dekusar et al 2014, Chernyakhovskaya 2016.…”

“…Traditionally, the LUFC is the average value of fuel costs per 1 kWh of electricity generated over the entire service life of a nuclear power plant. Since the costs associated with fuel are incurred throughout its life cycle, covering a period of up to 100 years or more, the method of reduced costs is used to take into account the disparity in monetary costs (OECD/NEA 1994, IAEA 2008, IAEA 2014; this actually takes into account the value of money.…”

“…Since the release of uranium takes place throughout the entire life cycle of a nuclear power plant, the additional income should be reduced (discounted) to a specific date in the same way as it is done when fuel costs are determined (OECD/NEA 1994, IAEA 2008, Dekusar et al 2014, Chernyakhovskaya 2016.…”

“…The calculations are based mainly on the physical and technical characteristics of a high-power power unit, adopted in accordance with (OECD/NEA 1994, Poplavsky et al 2010, Matveev and Khomyakov 2012, Alekseev et al 2016.…”

Determination of equilibrium fuel composition for fast reactor in closed fuel cycle

“…The developed algorithms and the results of this work are focused on the concept and design of high power reactors with sodium coolant, uranium-plutonium nitride fuel, high fuel burnup [1][2][3]. The increased diameter of the core and its significant flattening is typical for such reactors.…”