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Validation of TRACE Code Against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break

[+] Author Affiliations
Mian Xing, Yaodong Chen

State Nuclear Power Research Institute, SNPRI, Beijing, China

Xiao Hu

State Nuclear Power Technology Corporation Research & Development Center, SNPTRD, Beijing, China

Liangxing Li

Xi’an Jiaotong University, Xi’an, China

Weimin Ma

Royal Institute of Technology (KTH), Stockholm, Sweden

Paper No. ICONE21-15548, pp. V004T09A042; 11 pages
doi:10.1115/ICONE21-15548
From:
  • 2013 21st International Conference on Nuclear Engineering
  • Volume 4: Thermal Hydraulics
  • Chengdu, China, July 29–August 2, 2013
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5581-2
  • Copyright © 2013 by ASME

abstract

OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. ROSA-V Test 6-1 simulates a pressure vessel (PV) upper-head small break loss-of-coolant accident (SBLOCA) with a break size equivalent to 1.9% of the volumetrically scaled cross-sectional area of the reference PWR cold leg.

By building a TRACE calculation model of LSTF and PV upper-head, the paper dedicated to assess the effect of different modeling options and parameters on simulating thermal hydraulic behaviors of TRACE code. The results show that TRACE code well reproduce the physical phenomena involved in this type of SBLOCA scenarios. Almost all the events in the experiment are well predicted by the model based on TRACE code. In addition, the sensitivity of different models and parameters are investigated. For example, the code slightly overestimated the break mass flow from upper head which could affect the accuracy of the results significantly. The rising of core exit temperature (CET) is significantly influenced by the bypass flow area between downcomer and hot leg.

Copyright © 2013 by ASME

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