MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water

Bogoslovskaya, G. P.; Karpenko, A.A.; Kirillov, P. L.; Сорокин, А. П.

doi:10.1134/s0040601509030057

Therm. Eng.

2009

DOI: 10.1134/s0040601509030057

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MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water

G. P. Bogoslovskaya

A.A. Karpenko

P. L. Kirillov

et al.

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2012

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“…As a result of integrated studies, the code used for combined neutron-physical and thermohydraulic calculations of VVER-SKD was improved, the use of one-and two-loop working sections with supercritical water in a closed fuel cycle was analyzed, the thermal scheme chosen for a nuclear power plant with VVER-SKD-1600 was evaluated and optimized, the stress-strain state of a fuel element with oxide fuel was evaluated for the conditions of VVER-SKD operation, the normal and degraded heat-exchange regimes were studied, and the method and MIF-SKD code for calculating the thermohydraulic parameters of the reactor core taking account of different factors and parameter uncertainties were improved [10,11].…”

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Thermophysical studies for validating new-generation nuclear-reactor designs

Efanov¹,

Kalyakin²,

Сорокин³

2012

At Energy

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The results of thermophysical experimental and computational-theoretical studies for validating the enhanced efficiency and safety of new-generation nuclear reactors are presented. Critical heat flows have been studied to validate the design of fuel assemblies with elevated parameters on models with different spacer and intermediate lattices. Passive safety systems with a system of second-stage water tanks have been investigated. Heat transfer during coolant boiling in the core has been studied. Systems for monitoring hydrogen concentration and removal in nuclear power plants with VVER and an in-tank sodium purification system for maintaining an admissible impurity concentration in fast reactors have been tested.The present stage in the development of nuclear power is centered on perfecting nuclear power plants in order to improve their economic and operational efficiencies and to increase safety. One of the main avenues for solving this problem is to develop the inherent safety characteristics of nuclear power plants by combining safety properties with passive safety systems. The development and validation of these systems require a large series of thermophysical studies of nuclear power facilities in stationary and dynamic regimes, including the development of experimental models, experimental studies, and the development and improvement of numerical modeling methods.In recent years, specialists at the Physics and Power-Engineering Institute (FEI) have completed a large series of studies of nuclear power facilities with different coolants in stationary and dynamic regimes, including experimental and computational-theoretical thermophysical studies of nuclear power facilities in stationary, transient, and emergency regimes and the development and improvement of methods for numerical modeling of thermophysical processes. Most of the thermophysical studies were performed to validate the designs and safety of new-generation reactor facilities with liquid-metal coolants. A large volume of studies concerned the safety validation of the No. 4 unit with a BN-800 reactor being built at the Beloyarskaya nuclear power plant, the VVER-1000 reactor under design (AES-2006), and VVER with supercritical water.Increasing VVER Power Density and Efficiency. The following problems must be solved in order to increase the power density of VVER fuel assemblies: the admissible limits of fuel operation must be expanded and the working limits of the fuel must be brought closer to the design limits [1]. The first problem is closely related with improving fuel assembly design, specifically, using intermediate and spacer mixing lattices. The second problem is solved by computational-analytical methods by using better computer programs, eliminating excess conservatism, and other actions.Investigations of the thermohydraulic characteristics of 19-rod models of different VVER fuel assemblies show good (within 5%) agreement between the experimentally measured critical heat fluxes and those calculated using cross-section averaged models of the ...

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mentioning

confidence: 99%

Thermophysical studies for validating new-generation nuclear-reactor designs

Efanov¹,

Kalyakin²,

Сорокин³

2012

At Energy

Self Cite

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MIF-SKD computation code for thermohydraulic design of a reactor core cooled by supercritical-pressure water

Cited by 1 publication

References 1 publication

Thermophysical studies for validating new-generation nuclear-reactor designs

Thermophysical studies for validating new-generation nuclear-reactor designs

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