Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

被引：8

作者：

Dmitriev S.M. ^{[1
]}

Varentsov A.V. ^{[1
]}

Dobrov A.A. ^{[1
]}

Doronkov D.V. ^{[1
]}

Pronin A.N. ^{[1
]}

Sorokin V.D. ^{[1
]}

Khrobostov A.E. ^{[1
]}

机构：

[1] R. E. Alekseev Nizhnii Novgorod State Technical University, 24 Minin Str., Nizhnii Novgorod

来源：

Journal of Engineering Physics and Thermophysics | 2017年 / 90卷 / 4期

关键词：

coolant; fissile core; floating power plant; fuel assembly; spacer grid;

D O I：

10.1007/s10891-017-1641-7

中图分类号：

学科分类号：

摘要：

Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes). © 2017, Springer Science+Business Media, LLC.

引用

页码：941 / 950

页数：9

共 23 条

[21] EXPERIMENTAL INVESTIGATION OF CHANGES IN COOLANT FLOW-RATE THROUGH THE CORE IN THE CASE OF COOLANT PUMP FAILURE AND COMPARISON WITH RESULTS OF THEORETICAL INVESTIGATIONS ON A NUCLEAR-POWER PLANT WITH WWER-440 TYPE PRESSURIZED-WATER REACTOR
BARTOLOMEJ, GG
MERTINS, M
ZIMMER, E
KERNENERGIE, 1980, 23 (01): : 12 - 15
[22] Experimental investigations of heat transfer and temperature fields in models simulating fuel assemblies used in the core of a nuclear reactor with a liquid heavy-metal coolant
Belyaev I.A.
Genin L.G.
Krylov S.G.
Novikov A.O.
Razuvanov N.G.
Sviridov V.G.
Thermal Engineering, 2015, 62 (09) : 645 - 651
[23] The computational-and-experimental investigation into the head-flow characteristic of the two-stage ejector for the emergency core cooling system of the NPP with a water-moderated water-cooled power reactor
Parfenov Yu.V.
Thermal Engineering (English translation of Teploenergetika), 2013, 60 (09): : 660 - 662

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