Concept of an advanced power-generating unit with a BN-1200 sodium-cooled fast reactor

Rachkov, V. I.; Poplavskii, V. M.; Tsibulya, A. M.; Bagdasarov, Yu. E.; Vasiliev, B. A.; Kamanin, Yu. L.; Осипов, С. Л.; Kuzavkov, N.G.; Ershov, V. N.; Ashirmetov, M.R.

doi:10.1007/s10512-010-9286-z

Cited by 22 publications

(6 citation statements)

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“…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.…”

Section: Concept Of the Reactor Project And Criterion Functionals Conmentioning

confidence: 99%

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

et al. 2017

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Abstract. Technique of evaluation of multiplying and reactivity characteristics of fast reactor operating in the mode of multiple refueling is presented. We describe the calculation model of the vertical section of the reactor. Calculation validations of the possibility of correct application of methods and models are given. Results on the isotopic composition, mass feed, and changes in the reactivity of the reactor in closed fuel cycle are obtained. Recommendations for choosing perspective fuel compositions for further research are proposed.

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Section: Concept Of the Reactor Project And Criterion Functionals Conmentioning

confidence: 99%

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

et al. 2017

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“…To improve the reactor safety, given that radioactive sodium leakage had been taking place during operation of the BN-600 reactor primary circuit cold traps, a decision was made in the process of developing the large fast-neutron sodium-cooled reactor to install the purification system in the reactor tank (Rachkov et al 2010, Vasilyev et al 2016). The purification system's key components are the cold traps built in the reactor tank.…”

Section: Introductionmentioning

confidence: 99%

Conceptual issues of the cold filter trap development for the sodium coolant purification in fast-neutron reactors

Alekseev¹,

Kuzina²,

Сорокин³

2020

NUCET

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The paper presents the results of studying the peculiarities of heat and mass exchange in cold traps for the sodium purification of impurities in fast reactor circuits both in dedicated test areas simulating various trap components (isothermal sump, nonisothermal sump, filters, final cooling area) and in trap prototype models. As a result, a scientific rationale has been formed for developing traps of a unique design for various reactors. The impurity capacity of the traps is three to four times as high as that of the best foreign counterparts. Tests have shown these to be highly efficient in purifying sodium of oxygen and hydrogen and much less efficient in sodium purification of corrosion products and carbon. Taking into account the leakage of radioactive sodium during operation of the BN-600 reactor primary circuit traps, a decision was made to install the purification system in the reactor tank to improve the safety of the large fast reactor. It was resolved to exclude the accumulation of hydrogen in the primary circuit traps in nominal conditions. Two trap designs, with argon and sodium cooling, are discussed. It has been shown that operation of the reactor purification system with argon cooling will require 20 trap replacements during the reactor operating life and seven replacements if the deposition of hydrogen into the primary circuit cold traps is excluded. The sodium-cooled version of the trap built in the reactor tank has the same overall dimensions as the argon-cooled trap. The cooling sodium circulates in two trains: outside the jacketed working space body (up to 30% of the flow rate) and in the coil inside of the working space (up to 70% of the flow rate). Updates have been proposed to the trap design based on the calculations using the codes simulating the in-trap processes of heat and mass exchange.

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Experimental Research on Thermophysical Processes for Safety Validation of new-Generation Vver

et al. 2014

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The strategy for the development of nuclear power presupposes the construction of thermal water-cooled reactors and fast reactors [1][2][3]. At present, water moderated and cooled power reactors are the foundation of our country's nuclear power. The organizations of the State Corporation Rosatom have developed designs for new-generation reactor setups -AES-2006 and VVER-TOI, which are the result of evolutionary development and improvement of operating water moderated and cooled power reactors, which proved their reliability in thousands of reactor-years of accident-free operation. VVER are characterized by elevated power, the best economic and operational performance as well as enhanced safety compared with currently operating reactors. The thermal power is increased by, for example, using fuel assemblies with a new design incorporating heat-exchange intensifiers. The safety of the new designs of NPP with VVER is enhanced by means of the principle of technological diversity, which consists in combining active and passive safety systems. The passive systems ensure shutdown and prolonged removal of residual heat in the presence of a sealed loop or a depressurized loop and do not require operator intervention or an external long-time power source.To secure the possibility of increasing the power and validating the serviceability of new passive safety systems for VVER, a large-scale program of thermophysical and experimental studies has been completed.Increasing the Power Density and Efficiency of the VVER Core. The new NPP designs with VVER presuppose an improved design of the core as a whole and the fuel assemblies in particular. One promising avenue for increasing the capacity of power-generating units and the efficiency of the fuel cycle is to use fuel assemblies with improved thermohydraulic characteristics, which is accomplished by using mixing or intensifying lattices to increase heat transfer. Acting on the coolant flow these setups decrease the nonuniformity of the heating of the coolant (enthalpy) in the transverse section of a fuel assembly. They also make it possible to increase the turbulence properties of the flow, which causes moisture to settle on the walls of the fuel elements and promotes the removal of a large quantity of heat.In order to validate the heat-engineering reliability of a VVER core with fuel assemblies with mixing lattices, it is necessary to determine the effect of their design on the hydrodynamics and mass-transfer processes in the core. Important problems are the choice of the arrangement of the fuel assemblies and optimization of the design of the mixing lattices, which must possess the optimal combination of parameters such as the hydraulic resistance, the intensity of mixing of the coolant and a safety margin to crisis of boiling. The main method for studying the hydrodynamics and mass transfer in reactor fuel assemblies is the experimental investigation of fuel-assembly models on thermohydraulic stands: SVD-2, STF, and the TRASSER setup [4]. The main characteristics of the tripl...

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Concept of an advanced power-generating unit with a BN-1200 sodium-cooled fast reactor

Cited by 22 publications

References 1 publication

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

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

Conceptual issues of the cold filter trap development for the sodium coolant purification in fast-neutron reactors

Experimental Research on Thermophysical Processes for Safety Validation of new-Generation Vver

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