Thermal Hydraulic System Codes Performance in Simulating Buoyancy Flow Mixing Experiment in ROCOM Test Facility

被引:0
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作者
Coscarelli, Eugenio [1 ]
Lutsanych, Sergii [1 ]
D'Auria, Francesco [1 ]
机构
[1] Univ Pisa, San Piero Grado Nucl Res Grp GRNSPG, Via Livornese 1291, I-56122 Pisa, Italy
关键词
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中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
The MSLB (Main Steam Line Break) accident scenario is one of the severe abnormal transients that might occur in a NPP (Nuclear Power Plant). The main concerns of the MSLB are the potential return to power condition and the occurrence of PTS (Pressurized Thermal Shock) as a consequence of both rapid depressurization of the secondary circuit and the entrainment of cold water into the core region. Assessment of these issues is the main objective of integrated experimental tests carried out in the PKL-III and ROCOM facilities. The first test rig is aimed to simulate thermal-hydraulic phenomenology at the system level, whereas supporting ROCOM test facility is focused on the coolant mixing phenomenon that took place in the Reactor Pressure Vessel (RPV). Combination of these two typologies of experiments (integral effect test (IET) and separate effect test (SET)) provides appropriate experimental data for CFD and TH-SYS (Thermal Hydraulic-SYStem) codes validation against the relevant thermal hydraulic phenomena that occur during the MSLB. The main purpose of this study is to evaluate the capability of two TH-SYS codes TRACE V5 and CATHARE2 V2.5 to predict reasonably buoyancy driven mixing phenomena that affects the IVF (In-Vessel Flow) and the distribution of coolant temperature at the core inlet using 3-D porous media approach. Test 1.1 that had been carried out in ROCOM facility was selected to investigate the coolant mixing inside the RPV under flow conditions typical for a MSLB scenario. Averaging analysis of integral behaviour of the experimental and calculated temperature distributions inside the RPV has been performed.
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页数:10
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  • [1] ANALYSIS OF BUOYANCY-DRIVEN FLOW IN THE ROCOM TEST FACILITY
    Feng, Qingqing
    Bieder, Ulrich
    Hoehne, Thomas
    [J]. INTERNATIONAL YOUTH NUCLEAR CONGRESS 2016, IYNC2016, 2017, 127 : 44 - 53
  • [2] Experiments at the mixing test facility ROCOM for benchmarking of CFD codes
    Kliem, S.
    Suehnel, T.
    Rohde, U.
    Hoehne, T.
    Prasser, H-M
    Weiss, F-P
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2008, 238 (03) : 566 - 576
  • [3] Modeling of a buoyancy-driven flow experiment at the ROCOM test facility using the CFD codes CFX-5 and Trio_U
    Hoehne, Thomas
    Kliem, Soren
    Bieder, Ulrich
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2006, 236 (12) : 1309 - 1325
  • [4] Modeling of a buoyancy-driven flow experiment at the rocom test facility using the CFD-Code ansys Cfx
    Hoehne, Th.
    Kliem, S.
    Rohde, U.
    Weiss, F.-P.
    [J]. ATW-INTERNATIONAL JOURNAL FOR NUCLEAR POWER, 2007, 52 (03): : 168 - 174
  • [5] Application of the coupled code ATHLET-ANSYS CFX for the simulation of the flow mixing inside the ROCOM test facility
    Papukchiev, Angel
    Yang, Zhi
    [J]. PROGRESS IN NUCLEAR ENERGY, 2021, 137
  • [6] Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility
    Ding, Wen
    Zhang, Kui
    Zhang, Dalin
    Chen, Ronghua
    Ju, Zhongyun
    Xu, Caihong
    Wang, Ting
    Tian, Wenxi
    Qiu, Suizheng
    [J]. International Journal of Advanced Nuclear Reactor Design and Technology, 2023, 5 (03): : 115 - 122
  • [7] FLOW CONTROL IN A MIXING-VANE GRID TO ENHANCE THERMAL HYDRAULIC PERFORMANCE
    Ikeno, Tsutomu
    Sasakawa, Tatsuya
    Kakinoki, Shumpei
    Murase, Momonori
    [J]. PROCEEDINGS OF THE 18TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING 2010, VOL 4 PTS A AND B, 2011, : 215 - 222
  • [8] Modelling of multidimensional effects in thermal-hydraulic system codes under asymmetric flow conditions-Simulation of ROCOM tests 1.1 and 2.1 with ATHLET 3D-Module
    Pescador, E. Diaz
    Schaefer, F.
    Kliem, S.
    [J]. NUCLEAR ENGINEERING AND TECHNOLOGY, 2021, 53 (10) : 3182 - 3195
  • [9] Thermal hydraulic performance analysis of a post-CHF heat transfer test facility
    Liu, Qingqing
    Shi, Shanbin
    Sun, Xiaodong
    Kelly, Joseph
    [J]. NUCLEAR ENGINEERING AND DESIGN, 2018, 339 : 53 - 64
  • [10] Integral Test Facilities and Thermal-Hydraulic System Codes in Nuclear Safety Analysis
    Umminger, Klaus
    Del Nevo, Alessandro
    [J]. SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS, 2012, 2012