Fast Reactors and Accelerator Driven Systems Knowledge Base

IAEA-TECDOC--1157: LMFR core thermohydraulics: Status and prospects

Technical committee meeting on methods and codes for calculations of thermohydraulic parameters for fuel, absorber pins and assemblies of LMFR's with traditional and burner cores. Obninsk (Russian Federation) 27-31 Jul 1998

International Atomic Energy Agency, Vienna (Austria), IAEA-TECDOC--1157

Abstract

The Technical Committee Meeting (TCM) on “Methods and Codes for Calculations of Thermohydraulic Parameters for Fuel, Absorber Pins and Assemblies of LMFR’s with Traditional and Burner Cores” was held in Obninsk, Russian Federation from 27-31 July, 1998. This meeting was convened by the International Atomic Energy Agency (IAEA) and was hosted by the Federal Scientific Center, Institute of Physics and Power Engineering (IPPE). This meeting was organized as a forum for experts of Member States with fast reactor programmes to collectively review the present status and international progress made in LMFR’s core thermohydraulics.In addition, to exchange scientific and technical information, accumulated in different countries using their codes and methods of thermohydraulic calculation, to indicate the procedures of verification of the codes available for the calculation of reactor core thermal hydraulics under steady state and transient conditions and to discuss the issues requiring further research and development.

Attendance at this TCM included twenty seven participants and fifteen observers from seven countries (China, France, India, Japan, the Republic of Korea, the Russian Federation and the United Kingdom). Thirty papers were presented by the participants on a myriad of fast reactor core thermohydraulics including thermal-hydraulic subchannel analysis, thermal-hydraulic design of LMFR core, status of the validation of codes, analytical and experimental studies on local flow blockages in subassemblies, natural convection analysis and verification for LMFR assemblies, pre-boiling and boiling transients. Each presentation was followed by general discussion and the TCM concluded with a closing session on the evaluation of future code and method development requirements and exploration of areas for LMFR core thermohydraulics.

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Articles

Methods and codes for modeling thermohydraulic of fast reactor core subassemblies under nominal and non-nominal operating conditions
Sorokin, A.P.; Efanov, A.D.; Yuriev, Yu.S.; Zhukov, A.V.; Ushakov, P.A.; Bogoslovskaya, G.P. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

The status of studies on fast reactor core thermal hydraulics at PNC
Nishimura, M.; Ohshima, H.; Kamide, H.; Yamaguchi, K.; Yamaguchi, A. (O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, O-arai, Ibaraki-ken (Japan))

Thermal hydraulic design of PFBR core
Roychowdhury, D.G.; Vinayagam, P.P.; Ravichandar, S.C. (Indira Ghandi Centre for Atomic Research, Kalpakkam (India)) (and others)

Methodology for thermal hydraulic conceptual design and performance analysis of KALIMER core
Young-Gyun Kim; Won-Seok Kim; Young-Jin Kim; Chang-Kue Park (Korea Atomic Energy Research Institute, Taejon (Korea, Republic of))

Current status of thermohydraulic validation studies at CEA-Grenoble for the SIMMER-III code
Coste, P.; Pigny, S. (Commissariat a l'Energie Atomique (CEA), DRN/DTP/SMTH, Grenoble (France)); Meignen, R. (CISI Company (France))

3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers
Chvetsov, I.; Volkov, A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Low-Rynolds number k-epsilon turbulence model for calculation of fast-reactor-channel flows
Mikhin, V.I. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies
Sorokin, G.A.; Avdeev, E.F. (Institute of Nuclear and Power Engineering, Obninsk, Kaluga Region (Russian Federation)); Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors
Chvetsov, I.; Kouznetsov, I.; Volkov, A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

A simple analytical method to evaluate transient behaviours of metal fueled fast reactors
Ninokata, H.; Hizume, Y.; Sawada, T.; Endo, H. (Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan))

Codes for 3-dimensional thermohydraulic calculation of fast reactor core in steady state, transient and accident conditions
Buksha, Yu.K.; Marinenko, E.E.; Touzov, A.A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Calculation of fluid flow and heat transfer in a rod bundle with geometrical disturbance based on the 'locally exact' finite-difference scheme
Kriventsev, V.; Ninokata, H. (Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan))

Flow resistance in rod assemblies
Korsun, A.S.; Sokolova, M.S. (Moscow Engineering Physics Institute, Institute of Physics and Power Engineering, Moscow (Russian Federation))

Models and characteristics of interchannel exchange in pin bundles cooled by liquid metal
Bogoslovskaya, G.P.; Zhukov, A.V.; Sorokin, A.P. (State Scientific Center of Russian Federation, Obninsk, Kaluga Region (Russian Federation))

Development of subchannel analysis code MATRA-LMR for KALIMER subassembly thermal-hydraulics
Won-Seok Kim; Young-Gyun Kim (Korea Atomic Energy Research Institute, Taejon (Korea, Republic of))

Hydrodynamic module validation of THEHYCO-3DT code
Korsun, A.S.; Vitruk, S.G. (Moscow Engineering Physics Institute, Moscow (Russian Federation)); Ushakov, P.A.; Gabrianovich, B.N. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Special feature of temperature field formation in the system of interacting fuel subassemblies in fast reactor core
Bogatirjev, I.L.; Zhukov, A.V.; Mtjukhin, N.M.; Sorokin, A.P.; Ushakov, P.A. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Methods of statistical calculation of fast reactor core with account of influence of fuel assembly form change in process of campaign and other factors
Sorokin, G.A.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

The approximate thermal-model-testing method for non-stationary temperature fields in central zones of fast reactor assemblies
Mikhin, V.I.; Matukhin, N.M. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

The problems of thermohydraulics of prospective fast reactor concepts
Sedov, A.A. (RRC Kurchatov Institute, Moscow (Russian Federation))

The thermalhydraulics of a pin bundle with a helical wire wrap spacer. Modeling and qualification for a new sub-assembly concept
Valentin, B. (CEA/CEN Cadarache, DRN/DEC/SECA/LTEA, Saint-Paul-lez-Durances (France))

Thermal hydraulic behavior of sub-assembly local blockage in China experiment fast reactor
Yang Zhimin (China Institute of Atomic Energy, Beijing (China))

Design approach to local blockages
Roychowdhury, D.G.; Govindarajan, S.; Chetal, S.C.; Bhoje, S.B. (Indira Ghandi Centre for Atomic Research, Kalpakkam (India))

Numerical simulation of fuel assembly thermohydraulics of fast reactors with the partial blockage of cross section under the coolant
Zhukov, A.V.; Sorokin, A.P. (State Scientific Center of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

The questions of liquid metal two-phase flow modelling in the FBR core channels
Martsiniouk, D.Ye.; Sorokin, A.P. (State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

On development of analytical closure relationships for local wall friction, heat and mass transfer coefficients for sub-channel codes
Kornienko, Y. (State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Settlement substantiation of the passive devices shutdown fast reactors by trip the absorbing rod in case of anticipated accident
Portianoy, A.G.; Serdun, E.N.; Sorokin, A.P.; Uhov, V.A.; Egorov, V.S. (State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Modelling problems of emergency natural convection heat removal in the upper plenum of fast reactors using water
Ushakov, P.A.; Sorokin, A.P. (State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))

Fast reactor core thermal-hydraulic analyses during transition from forced to natural circulation
Nishimura, M.; Ohshima, H.; Kamide, H. (O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, O-arai, Ibaraki-ken (Japan))

Natural convection as the way of heat removal from fast reactor core at cooldown regimes
Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A. (State Scientific Centre of Russian Federation, Institute of Physics and Power Engineering, Obninsk, Kaluga Region (Russian Federation))