Loss-of-heat-sink transient simulation with RELAP5/Mod3.3 code for the ATHENA facility

被引:0
|
作者
Del Moro, T. [1 ]
Giannetti, F. [1 ]
Puviani, P. Cioli [2 ]
Di Piazza, I. [3 ]
Diamanti, D. [3 ]
Tarantino, M. [3 ]
机构
[1] Sapienza Univ Rome, Corso Vittorio Emanuele II 244, I-00186 Rome, Italy
[2] Politecn Torino, Corso Duca degli Abruzzi 24, I-10129 Turin, Italy
[3] ENEA, I-40032 Bologna, Italy
关键词
LFR; ALFRED; ATHENA; RELAP5/Mod3.3; LOHS;
D O I
10.1016/j.anucene.2024.110948
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
ATHENA (Advanced Thermal-Hydraulic Experiment for Nuclear Applications) is a large multipurpose pool-type lead-cooled facility under construction at the Mioveni site in Romania. It has been identified by the FALCON (Fostering ALfred CONstruction) Consortium to characterize large to full-scale ALFRED components, to conduct integral tests, and to investigate the main thermal-hydraulic phenomena inherent in pool-type systems. ATHENA is representative of ALFRED in terms of the difference in height of the thermal barycenters of the heat source and heat sink, i.e., 3.3 m, in order to reproduce the buoyancy forces in the system. Similar to ALFRED's design, ATHENA minimizes thermal stratification within the main vessel even under natural circulation conditions, through an internal structure referred to as "barrel". This structure directs the fluid flow towards the main vessel, preventing fluid stagnation near the vessel itself. The paper initially provides a steady-state thermal-hydraulic characterization of the facility, including details of the numerical model developed using the RELAP5/Mod3.3 thermal-hydraulic code. Then, focus is given to the transient analysis considering as a reference scenario a Loss-of-Heat-Sink (LOHS) accidental transient. In this scenario, the Main Circulation Pump (MCP) is assumed to remain operational while the Core Simulator (CS) is deactivated once the lead temperature at the Main Heat Exchanger (MHX) outlet reaches a predefined threshold. A sensitivity analysis is conducted with set points of 430 degrees C, 450 degrees C, 470 degrees C, and 490 degrees C, assessing the system's response following MHX isolation from the secondary loop. The study evaluates the impact of different CS deactivation set points on reactor SCRAM delay (reducing CS power to a level representative of decay heat) as well as on system maximum and minimum temperatures.
引用
收藏
页数:13
相关论文
共 50 条
  • [1] Validation of RELAP5 MOD3.3 code for entrainment phenomenon against FATE test facility
    Hu, Xiao
    Zhang, Peng
    Zhang, Lei
    Li, Wei
    Xing, Mian
    Chen, Lian
    Chen, Peipei
    Chang, Huajian
    Chen, Renzong
    PROGRESS IN NUCLEAR ENERGY, 2018, 106 : 425 - 432
  • [2] Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3
    Wu, Pan
    Xiong, Xiaofei
    Shan, Jianqiang
    Gou, Junli
    Zhang, Bin
    Zhang, Bo
    NUCLEAR ENGINEERING AND DESIGN, 2016, 310 : 418 - 428
  • [3] 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
  • [4] 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,
  • [5] RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation
    Prosek, Andrej
    Matkovic, Marko
    SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS, 2018, 2018
  • [6] Analysis of a loss of residual heat removal system during mid-loop conditions at PKL facility using RELAP5/Mod3.3
    Carlos, S.
    Villanueva, J. F.
    Martorell, S.
    Serradell, V.
    NUCLEAR ENGINEERING AND DESIGN, 2008, 238 (10) : 2561 - 2567
  • [7] Assessment of RELAP5/MOD3.3 Against Single Rod Reflooding Experiments
    Lymperea, N.
    Nikoglou, A.
    Hinis, E.
    ASCE-ASME JOURNAL OF RISK AND UNCERTAINTY IN ENGINEERING SYSTEMS PART B-MECHANICAL ENGINEERING, 2018, 4 (03):
  • [8] Modeling of droplet diameter changes during reflood into RELAP5/Mod3.3
    Park, Ju Hyun
    Jeong, Yong Hoon
    NUCLEAR ENGINEERING AND DESIGN, 2022, 393
  • [9] Implementation of Lead-Lithium as working fluid in RELAP5/Mod3.3
    Barone, Gianluca
    Martelli, Daniele
    Forgione, Nicola
    FUSION ENGINEERING AND DESIGN, 2019, 146 : 1308 - 1312
  • [10] RELAP5/MOD3.3 Best Estimate Analyses for Human Reliability Analysis
    Prosek, Andrej
    Mavko, Borut
    SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS, 2010, 2010