Validation of RELAP5/MOD3.3 for subcooled boiling, flashing and condensation in a vertical annulus

被引:22
|
作者
Fullmer, William D. [1 ]
Kumar, Vineet [2 ]
Brooks, Caleb S. [2 ]
机构
[1] Univ Colorado, Dept Chem & Biol Engn, Boulder, CO 80309 USA
[2] Univ Illinois, Dept Nucl Plasma & Radiol Engn, 104 South Wright St, Urbana, IL 61801 USA
关键词
RELAP5; Subcooled boiling; Flashing; Condensation; Validation; 2-PHASE FLOW; VOID FRACTION; HEAT-TRANSFER; START-UP; PRESSURE; INSTABILITIES; WATER; SYSTEMS; CHANNEL; STEAM;
D O I
10.1016/j.pnucene.2016.08.013
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
Continued development of system analysis codes has resulted in the recovery of conservatisms originally imposed on nuclear power reactors, allowing for an increase in the capacity of commercial nuclear reactors. These codes also play an instrumental role in the design and certification of new reactor systems. With the increased demand for passive natural circulation and gravity driven cooling options, these codes are met with the new challenge of simulating low pressure, low flow conditions. The objective of this work is to demonstrate the effectiveness of the widely used RELAP5/MOD3.3 code to simulate boiling, condensing and flashing flows under such conditions. Two-phase flow data in an internally heated vertical annulus with inner diameter of 19.1 mm and outer diameter of 38.1 mm is utilized for validation of the RELAP5/MOD3.3. The code calculation of pressure, temperature, void fraction, interfacial area concentration, and void weighted gas velocity along the 4.5 m test section is compared with data at five axial locations. In the 2.8 m heated section of the channel the code predictions compare favorably in general, although the error does increase at low system pressure. Beyond the heated length, code predictions of condensation and flashing show more noticeable disagreement along the 1.7 m unheated section. Condensation is consistently under-predicted. Flashing varies from relatively good agreement to complete failure, depending on the conditions at the exit of the heated section. User options related to boiling and condensation are also assessed, and shown to have marginal improvements in some conditions. In general the code consistently predicts the point of net vapor generation too soon along the heated length at low mass flux, over predicts the void fraction at the end of the heated length, and has large scatter in void fraction agreement at the end of the channel. (C) 2016 Elsevier Ltd. All rights reserved.
引用
收藏
页码:205 / 217
页数:13
相关论文
共 50 条
  • [31] Pre and Post Test Analysis of LBLOCA Late Reflood Phase in ATLAS Using RELAP5/MOD3.3
    Lee, Seok Ho
    Kim, Han Gon
    ICONE16: PROCEEDING OF THE 16TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2008, VOL 3, 2008, : 275 - 282
  • [32] RELAP5/MOD3.3 Simulation of LOFT LP-FW-1 Total Loss of Feedwater Test
    Prosek, Andrej
    29TH INTERNATIONAL CONFERENCE NUCLEAR ENERGY FOR NEW EUROPE (NENE 2020), 2020,
  • [33] A study on low-pressure subcooled flow boiling using RELAP5/MOD3.4
    Xu, Caihong
    INTERNATIONAL YOUTH NUCLEAR CONGRESS 2016, IYNC2016, 2017, 127 : 387 - 397
  • [34] Validation of RELAP5 MOD3.3 for 1% reactor pressure vessel top head break loss of coolant accident via the ATLAS facility
    Yoon, Ho Joon
    Alyammahi, Nourah
    Al-Yahia, Omar S.
    Leung, Raymond
    NUCLEAR ENGINEERING AND DESIGN, 2022, 400
  • [35] Analysis of Channel Blockage of MNSR Reactor Using the System Thermal-Hydraulic Code RELAP5/MOD3.3
    Adu, Simon
    Nyarko, B. J. Benjamin
    Emi-Reynolds, Geoffrey
    Darko, Emmanuel O.
    Horvatovic, Ivan
    Menzel, Francine
    D'Auria, Francesco
    24TH INTERNATIONAL CONFERENCE NUCLEAR ENERGY FOR NEW EUROPE, (NENE 2015), 2015,
  • [36] Development and verification of tritium transport code based on RELAP5/ MOD3.3 with generic model towards COOL blanket
    Wang, Wenjia
    Zhao, Xueli
    Cheng, Xiaoman
    Lu, Shuailing
    Liu, Songlin
    FUSION ENGINEERING AND DESIGN, 2024, 199
  • [37] LOCA plus Loss of One Emergency Core Cooling System Simulated by RELAP5/MOD3.3 Patch 05
    Prosek, Andrej
    30TH INTERNATIONAL CONFERENCE NUCLEAR ENERGY FOR NEW EUROPE (NENE 2021), 2021,
  • [38] ANALYSIS OF THE ISP-50 DIRECT VESSEL INJECTION SBLOCA IN THE ATLAS FACILITY WITH THE RELAP5/MOD3.3 CODE
    Sharabi, Medhat
    Freixa, Jordi
    NUCLEAR ENGINEERING AND TECHNOLOGY, 2012, 44 (07) : 709 - 718
  • [39] On the subcooled critical flow model in RELAP5/MOD3
    Yeung, WS
    Shirkov, J
    NUCLEAR TECHNOLOGY, 1996, 114 (01) : 141 - 145
  • [40] Improvements of RELAP5/Mod3.3 heat transfer capabilities for simulation of in-pool passive power removal systems
    Narcisi, Vincenzo
    Melchiorri, Lorenzo
    Giannetti, Fabio
    ANNALS OF NUCLEAR ENERGY, 2021, 160