MECHANISTIC PREDICTION OF CHF IN FUEL ASSEMBLY USING THE SUBCHANNEL CODE CAPE

被引:0
作者
Yoshida, Ryo [1 ]
Yoshida, Kenji [1 ]
Kataoka, Isao [1 ]
Naito, Masanori
机构
[1] Osaka Univ, Dept Mech Engn, Suita, Osaka 5650871, Japan
来源
ICONE 17: PROCEEDINGS OF THE 17TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 3 | 2009年
关键词
D O I
暂无
中图分类号
TH [机械、仪表工业];
学科分类号
0802 ;
摘要
In order to predict the critical power or void fraction in BWR fuel bundles and the DNB heat flux of PWR fuel assemblies, the boiling transition analysis code called "CAPE" with mechanistic models has been developed in the IMPACT project by NUPEC The objective of the CAPE code development is to perform with good accuracy the safety evaluation for a new type or improved fuel bundle design of BWR and PWR without full-scale experiments or any tuning parameters in the analysis code. In the present study dryout heat fluxes of BWR fuel assembly were analyzed by the CAPE code and compared with experimental data of BFBT benchmark test carried out by NUPEC In the CAPE code, mechanistic model of liquid film dryout in annular flow is used considering entrainment and deposition of droplet In such mechanistic prediction of dryout, the correlations of entrainment and deposition rates play quite Important roles and many correlations have been developed. In addition to the original correlations in the CAPE code, several typical correlations, which are widely used in the analysis of annular dispersed flow, were tested for the prediction of dryout. The results indicated that the CAPE code satisfactorily predicted dryout heat fluxes of fuel assembly for wide range of pressure, mass flux, subcooling and bundle geometries obtained in BFBT benchmark test. The accuracy of prediction depends upon the combination of correlations of entrainment and deposition rates. The evaluation of correlations of entrainment and deposition rates was carried out.
引用
收藏
页码:69 / 76
页数:8
相关论文
共 4 条
  • [1] VOID FRACTION DISTRIBUTION IN BWR FUEL ASSEMBLY AND EVALUATION OF SUBCHANNEL CODE
    INOUE, A
    KUROSU, T
    AOKI, T
    YAGI, M
    MITSUTAKE, T
    MOROOKA, S
    [J]. JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 1995, 32 (07) : 629 - 640
  • [2] *JAP SOC MULT FLOW, 1994, ZJ956594001 PNC JAP
  • [3] Study on analytical prediction of forced convective CHF based on multi-fluid model
    Kataoka, I
    Kodama, S
    Tomiyama, A
    Serizawa, A
    [J]. NUCLEAR ENGINEERING AND DESIGN, 1997, 175 (1-2) : 107 - 117
  • [4] KATAOKA I., 1996, JAPANESE J MULTIPHAS, V10, P171, DOI [https://doi.org/10.3811/jjmf.10.171, DOI 10.3811/JJMF.10.171]