Occurrence of thermal stratification in sodium cooled fast reactor piping

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved in the HLM technology development. In particular, several experimental campaign employing HLM loop and pool facilities (CIRCE [1], NACIE [2], HELENA [3], HERO [4]) are carried out in order to support HLM technologies development. In this frame, suitable experiments were carried out on the CIRCE pool facility refurbished with the Integral Circulation Experiment (ICE) test section in order to investigate the thermal hydraulics and the heat transfer in grid spaced Fuel Pin Bundle cooled by liquid metal providing, among the others aim, experimental data in support of codes validation for the European fast reactor development. The study of thermal stratification in large pool reactor is relevant in the design of HLM nuclear reactor especially for safety issue. Thermal stratification should induce thermomechanical stresses on the structures and in accidental scenario conditions, could opposes to the establishment of natural circulation which is a fundamental aspect for the achievements of safety goals required in the GEN-IV roadmap. In the present work, a Protected Loss of Heat Sink with Loss Of Flow (PLOHS+LOF) scenario is experimentally simulated and the mixed convection with thermal stratification phenomena is investigated during the simulated transient, as foreseen in the frame of Horizon 2020 SESAME project [5]. A full characterization of thermal stratification inside the pool is presented, and the main results gained during the run are reported. The two tests named A (20 h) and B (6 h) here reported, essentially differs for the power supplied to the fuel bundle during the full power run (800 kW and 600 kW respectively). After the transition to natural circulation conditions, the power supplied to the bundle is decreased to about 30 kW simulating the decay heat. Finally the Nusselt number for the central subchannel of the fuel bundle simulator (FPS) is evaluated and compared with values obtained from Ushakov and Mikityuk correlations [6–7].

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Experimental Study on Thermal Stratification in a Reactor Vessel of Innovative Sodium-Cooled Fast Reactor—Mitigation Approach of Temperature Gradient across Stratification Interface—

Journal of Nuclear Science and Technology ◽

10.1080/18811248.2010.9711659 ◽

2010 ◽

Vol 47 (9) ◽

pp. 829-838 ◽

Cited By ~ 4

Author(s):

Nobuyuki KIMURA ◽

Hiroyuki MIYAKOSHI ◽

Hideki KAMIDE

Keyword(s):

Experimental Study ◽

Temperature Gradient ◽

Fast Reactor ◽

Thermal Stratification ◽

Reactor Vessel

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Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor

Science and Technology of Nuclear Installations ◽

10.1155/2012/351809 ◽

2012 ◽

Vol 2012 ◽

pp. 1-13 ◽

Cited By ~ 1

Author(s):

Toshikazu Takeda ◽

W. F. G. van Rooijen ◽

Katsuhisa Yamaguchi ◽

Masayoshi Uno ◽

Yuji Arita ◽

...

Keyword(s):

Fast Reactor ◽

Thermal Stratification ◽

Low Flow ◽

Detailed Calculation ◽

Thermal Hydraulics ◽

New Methods ◽

Materials Modeling ◽

Fuel Assemblies ◽

Intermediate Heat Exchanger ◽

Design Calculations

This paper discusses the objectives and results of a multiyear R&D project to improve the modeling accuracy for the detailed calculation of the Japanese Sodium-cooled Fast Reactor (JSFR), although the preliminary design of JSFR is prepared using conventional methods. For detailed design calculations, new methods are required because the JSFR has special features, which cannot be accurately modeled with existing codes. An example is the presence of an inner duct in the fuel assemblies. Therefore, we have developed new calculational and experimental methods in three areas: (1) for neutronics, we discuss the development of methods and codes to model advanced FBR fuel subassemblies, (2) for fuel materials, modeling and measurement of the thermal conductivity of annular fuel is discussed, and (3) for thermal hydraulics, we describe advances in modeling and calculational models for the intermediate heat exchanger and the calculational treatment of thermal stratification in the hot plenum of an FBR under low flow conditions. The new methods are discussed and the verification tests are described. In the validation test, measured data from the prototype FBR Monju is partly used.

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