A validation of ATR LOCA thermal-hydraulic code with a statistical approach

被引:0
|
作者
Mochizuki, H [1 ]
机构
[1] Japan Nucl Cycle Dev Inst, Tokai, Ibaraki 3191184, Japan
关键词
statistical method; loss of coolant; downcomer break; evaluation model; best estimate model; ATR reactor; fuel element clusters; fuel bundles; cladding; fuel cans; temperature dependence;
D O I
10.1080/18811248.2000.9714946
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
When cladding temperatures are measured for a blowdown experiment, cladding temperatures at the same elevation in the fuel bundle have usually some differences due to eccentricity of the fuel bundle and other reasons such as biased two-phase flow. In the present paper, manufacturing tolerances and uncertainties of thermal-hydraulics are incorporated into a LOCA code that is applied with the statistical method. The present method was validated with the results of different blowdown experiments conducted using the 6 MW blowdown facility simulating the Advanced Thermal Reactor (ATR). In the present statistical method, the code was modified to run fast in order to calculate the blowdown thermal-hydraulics a lot of times with the code using different sets of input data. These input data for sizes and empirical correlations are prepared by the effective Monte-Carlo method based on the distribution functions deduced by the measured manufacturing errors and the uncertainties of thermal hydraulics. The calculated curves express uncertainties due to the different input deck. The uncertainty band and tendency of the cladding temperature were dependent on the beak sizes in the experiment. The measured results were traced by the present method.
引用
收藏
页码:697 / 709
页数:13
相关论文
共 50 条
  • [41] Development of a thermal-hydraulic analysis code for annular fuel assemblies
    Vishnoi, A. K.
    Chandraker, D. K.
    Vijayan, P. K.
    KERNTECHNIK, 2012, 77 (01) : 12 - 17
  • [42] Implementation of CFD module in the KORSAR thermal-hydraulic system code
    Yudov, Yu. V.
    Danilov, I. G.
    Chepilko, S. S.
    KERNTECHNIK, 2015, 80 (04) : 359 - 365
  • [43] TUF - 2-FLUID CODE FOR THERMAL-HYDRAULIC ANALYSIS
    LIU, W
    YOUSEF, W
    PASCOE, J
    TOMASONE, A
    WILLIAMS, M
    LUXAT, JC
    10TH ANNUAL CONFERENCE 1989 - CANADIAN NUCLEAR SOCIETY, PROCEEDINGS VOL 3, 1989, : C1 - C9
  • [44] Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies
    Castellanos-Gonzalez, Duvan A.
    Losada Moreira, Joao Manoel
    Maiorino, Jose Rubens
    Carajilescov, Pedro
    SCIENCE AND TECHNOLOGY OF NUCLEAR INSTALLATIONS, 2018, 2018
  • [45] THINC-IV, A NEW THERMAL-HYDRAULIC CODE FOR PWR THERMAL DESIGN
    CHU, PT
    HOCHREITER, LE
    CHELEMER, H
    BOMAN, LH
    TONG, LS
    TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY, 1972, 15 (02): : 876 - +
  • [46] Development of a thermal-hydraulic system code for simulators based on RELAP5 code
    Lin, M
    Su, Y
    Hu, R
    Zhang, RH
    Yang, YH
    NUCLEAR ENGINEERING AND DESIGN, 2005, 235 (06) : 675 - 686
  • [47] Validation of the thermal-hydraulic system code ATHLET based on selected pressure drop and void fraction BFBT tests
    Di Marcello, Valentino
    Escalante, Javier Jimenez
    Espinoza, Victor Sanchez
    NUCLEAR ENGINEERING AND DESIGN, 2015, 288 : 183 - 194
  • [48] Lead coolant modeling in system thermal-hydraulic code HYDRA-IBRAE/LM and some validation results
    Mosunova, N. A.
    Alipchenkov, V. M.
    Pribaturin, N. A.
    Strizhov, V. F.
    Usov, E., V
    Lobanov, P. D.
    Afremov, D. A.
    Semchenkov, A. A.
    Larin, I. A.
    NUCLEAR ENGINEERING AND DESIGN, 2020, 359
  • [49] Thermal-hydraulic validation of two-phase models in THUNDER code against benchmark results and CFD codes
    Castellanos-Gonzalez, Duvan A.
    Maiorino, Jose Rubens
    Monteiro, Deiglys Borges
    Losada Moreira, Joao Manoel
    NUCLEAR ENGINEERING AND DESIGN, 2020, 369 (369)
  • [50] Code Development and Validation for Thermal-hydraulic Analysis of Helical Coil Steam Generator Based on RELAP5
    Lian Q.
    Tian W.
    Qiu S.
    Su G.
    Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2019, 53 (06): : 1007 - 1013