EFFECT OF MODELLING CROSS-FLOW ON THE SIMULATION OF SBLOCA EXPERIMENTS OF PSB-VVER TEST FACILITY

Yıl 2019, Cilt: 31 Sayı: 2, 105 - 117, 31.12.2019

Tuğçe Türkal Şule Ergün

Öz

In this work, small break loss of coolant accident (SBLOCA) was modelled at PSB-VVER test facility. By using a RELAP5 input deck prepared during OECD PSB-VVER project, cross-flow flow between the thermalhydraulic channels is modeled. As the result of modelling lateral flow, the realistic situations where the flow is mixed the results of simulations were investigated. The results of simulations were compared with the results obtained by simulating the cases where flow mixing was not modeled and effect of modelling cross-flow on the calculation of peak clad temperature, actuation time of the accumulator and the low pressure emergency core cooling system in addition to mass flow rate was observed. The result of modelling cross-flow was compared with the simulation results for which the cross-flow effect was ignored. Code validation was performed by comparing PSB-VVER test facility experimental results with the simulation results obtained by changing the core nodalization in RELAP5 MOD 3.5 input file for cross-flow modelling.

Anahtar Kelimeler

PSB-VVER, VVER, RELAP5, loss of coolant accident

Kaynakça

• [1] O.I. Melikhov, I.V. Elkin, I.A. Lipatov, A.V. Kapustin, S.M. Nikonov, G.I. Dremin, Pre-test Experiment Proposal Report (Test 3), OECD PSB-VVER Project, PSB-VVER Report, PSB-16, EREC, Electrogorsk, Russia, 2003
• [2] O.I. Melikhov, I.V. Elkin, I.A. Lipatov, A.V. Kapustin, S.M. Nikonov, G.I. Dremin, S.A. Galchanskaya, A.A. Rovnov, V.I. Gudkov, A.I. Antonova, A.F. Chalych, Post-test Quick Look Report (Test 3), OECD PSB-VVER Project, PSB-VVER Report , PSB-17, EREC, Electrogorsk, Russia, 2004
• [3] O.I. Melikhov, I.V. Elkin, I.A. Lipatov, A.V. Kapustin, S.M. Nikonov, G.I. Dremin, A.A. Rovnov, V.I. Gudkov, A.I. Antonova, Post-test Full Experimental Data Report (Test 3), OECD PSB-VVER Project, PSB-VVER Report, PSB-18, EREC, Electrogorsk, Russia, 2004
• [4] O.I. Melikhov, I.V. Elkin, I.A. Lipatov, A.V. Kapustin, S.M. Nikonov, G.I. Dremin, A.A. Rovnov, V.I. Gudkov, A.I. Antonova, Standard Problem Definition Report (Test 3), OECD PSB-VVER Project, PSB-VVER Report, PSB-21, EREC, Electrogorsk, Russia, 2004
• [5] O.I. Melikhov, I.V. Elkin, I.A. Lipatov, A.V. Kapustin, G.I. Dremin, Standard Problem Final Analysis Report (Test 3), OECD PSB-VVER Project, PSB-VVER Report, PSB-22, EREC, Electrogorsk, Russia, 2004
• [6] P.A.L Reis, A.L. Costa, C. Pereira, C.A.M. Silva, M.A.F. Veloso, A.Z. Mesquita, Sensitivity Analysis to a RELAP5 Nodalization Developed for a Typical TRIGA Research Reactor, Brasil, 2011
• [7] L.R.M. Mól, P.A.L Reis, A.L. Costa, C. Pereira, C.A.M. Silva, M.A.F. Veloso, A.Z. Mesquita, Thermal-Hydraulic Sensitivity Analysis of an IPR-R1 TRIGA Reactor Modeling in The RELAP5 Code, Brasil, 2011
• [8] N. Todreas, M.S. Kazimi, Nuclear Systems II Elements of Thermal Hydraulic Design, Massachusetts Institute of Technology, USA, 1990