0
Research Papers

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

[+] Author and Article Information
Yao Xiao

Xi’an Jiaotong University,
Xi’an, Shaanxi, 710049, China;
Massachusetts Institute of Technology,
Cambridge, MA 02139
e-mail: yaoyuan.xiao@gmail.com

Lin-Wen Hu

Massachusetts Institute of Technology,
Cambridge, MA 02139
e-mail: lwhu@mit.edu

Suizheng Qiu

Xi’an Jiaotong University,
Xi’an, Shaanxi 710049, China
e-mail: szqiu@mai.xjtu.ed.cn

Dalin Zhang

Xi’an Jiaotong University,
Xi’an, Shaanxi 710049, China
e-mail: dlzhang@mai.xjtu.ed.cn

Su Guanghui

Xi’an Jiaotong University,
Xi’an, Shaanxi 710049, China
e-mail: ghsu@mai.xjtu.ed.cn

Wenxi Tian

Xi’an Jiaotong University,
Xi’an, Shaanxi 710049, China
e-mail: wxtian@mail.xjtu.ed.cn

1Corresponding author.

Manuscript received August 2, 2014; final manuscript received September 27, 2014; published online February 9, 2015. Assoc. Editor: Dmitry Paramonov.

ASME J of Nuclear Rad Sci 1(1), 011007 (Feb 09, 2015) (7 pages) Paper No: NERS-14-1031; doi: 10.1115/1.4026394 History: Received August 02, 2014; Accepted November 14, 2014; Online February 09, 2015

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

FIGURES IN THIS ARTICLE
<>
Copyright © 2015 by ASME
Your Session has timed out. Please sign back in to continue.

References

Forsberg, C. W., Hu, L. W., Peterson, P. F., and Allen, T., 2012, “Fluoride-Salt-Cooled High-Temperature Reactors (FHRs) for Power and Process Heat,” Report MIT-ANP-TR-143, Massachusetts Institute of Technology, Cambridge, MA.
SINAP, 2012, “Pre-Conceptual Design of a 2 MW Pebble-Bed Fluoride Salt Coolant High Temperature Test Reactor,” Shanghai Institute of Applied Physics, Shanghai, China.
Jiang, M. H., Xu, H. J., and Dai, Z. M., 2012, “Advanced Fission Energy Program-TMSR Nuclear Energy System,” Bull. Chin. Acad. Sci., 27(3), pp. 366–374.
Dai, Z. M., 2014, “Thorium Molten Salt Reactor System,” Shanghai Institute of Applied Physics, Shanghai, China (in Chinese).
Bickel, J. E., Laufer, M. R., Li, L., Cisneros, A. T., and Peterson, P. F., 2010, “Conceptual Design, Experiments, and Analysis for the Core of an FHR-16 Test Reactor,” Proceedings of the International Congress on Advances in Nuclear Power Plants 2010, ICAPP 2010, June 13–June 17, American Nuclear Society, pp. 1281–1291.
SINAP, 2012, “TMSR Internal Technical Report XDA02010200-TL-2012-09,” Shanghai Institute of Applied Physics, Shanghai, China (in Chinese).
Xiao, Y., Hu, L. W., Forsberg, C., Qiu, S. Z., and Su, G. H., 2013, “Licensing Considerations of a Fluoride Salt Cooled High Temperature Test Reactor,” Proceedings of the 21st International Conference on Nuclear Engineering, Chengdu, China, July 29–Aug. 2.
Xiao, Y., Hu, L. W., Forsberg, C., Qiu, S. Z., Su, G. H., Chen, K., and Wang, N. X., 2014, “Analysis of the Limiting Safety System Settings of a Fluoride Salt Cooled High Temperature Test Reactor,” Nucl. Technol., 187(3), pp. 221–234. 10.13182/NT13-93
Collier, J. G., and Thome, J. R., 1994, Convective Boiling and Condensation, Oxford University Press, New York.
Yujun, G., Jinling, Z., Suizheng, Q., Guanghui, S., Dounan, J., and Zhenwan, Y., 1997, “MITARS: A Thermal Hydraulic Analysis Code for Nuclear Reactor System,” Proceedings of the 8th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Kyoto, Japan, Sept. 30–Oct. 4, Atomic Energy Society of Japan, pp. 1212–1219.
Su, G. H., Jia, D. N., Fukuda, K., and Guo, Y. J., 2001, “Theoretical Study on Density Wave Oscillation of Two-Phase Natural Circulation Under Low Quality Conditions,” J. Nucl. Sci. Technol., 38(8), pp. 607–613. 10.1080/18811248.2001.9715073
Gear, C. W., 1971, Numerical Initial Value Problems in Ordinary Differential Equation, Prentice-Hall, Upper Saddle River, NJ.
Hindmarsh, A. C., 1974, GEAR: Ordinary Differential Equation System Solver, Lawrence Livermore Laboratory, Livermore, CA.
Zuying, G., and Lei, S., 2002, “Thermal Hydraulic Calculation of the HTR-10 for the Initial and Equilibrium Core,” Nucl. Eng. Des., 218(1–3), pp. 51–64. 10.1016/S0029-5493(02)00198-X
Wakao, N., Kaguei, S., and Funazkri, T., 1979, “Effect of Fluid Dispersion Coefficients on Particle-to-Fluid Heat-Transfer Coefficients in Packed-Beds - Correlation of Nusselt Numbers,” Chem. Eng. Sci., 34(3), pp. 325–336. 10.1016/0009-2509(79)85064-2
Wakao N., and Kaguei, S., 1982, Heat and Mass Transfer in Packed Beds[M], Gordon and Breach, New York.
Forsberg, C. W., Peterson, P. F., and Williams, D. F., 2005, “Liquid-Salt Cooling for Advanced High-Temperature Reactors,” Proceedingsof the American Nuclear Society—International Congress on Advances in Nuclear Power Plants 2005, ICAPP’05, Seoul, Korea, May 15–19, American Nuclear Society, pp. 2080–2095.
Kim, S. J., Hu, L. W., and Dunn, F., 2013, “Thermal-Hydraulic Analysis for High Enrichment Uranium (HEU) and Low Enrichment Uranium (LEU) Transitional Core Conversion of the MIT Research Reactor,” Nucl. Technol., 182(3), pp. 315–334.
MITR-Staff, 2011, “Safety Analysis Report for the MIT Research Reactor,” MIT Nuclear Reactor Laboratory, Cambridge, MA.
SINAP, 2013, “Current Status of the TMSR Project in China,” Shanghai Institute of Applied Physics, Shanghai, China.
Song, Y.-M., Ma, Y.-L., and Zhou, Z.-W., 2010, “Real-Time Simulation of Neutron Space-Time Kinetics for High-Temperature Gas-Cooled Reactor,” Yuanzineng Kexue Jishu/Atom. Energy Sci. Technol., 44(2), pp. 188–192.
Ingersoll, D. T., Forsberg, C. W., Ott, L. J., Williams, D. F., et al. , 2004, “Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR),” , Oak Ridge National Laboratory.
HAYNES, 2002, “HASTELLOY® N alloy,” Haynes International, Inc., Kokomo, IN.
NUREG/CR-6844, 2004, “TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables (PIRTs) for Fission Product Transport Due to Manufacturing, Operations, and Accidents,” Office of Nuclear Reactor Regulation, Nuclear Regulation Commission, Washington, DC.

Figures

Grahic Jump Location
Fig. 2

Model of TMSR-SF fuel pebble

Grahic Jump Location
Fig. 1

TMSR-SF cross-sectional view

Grahic Jump Location
Fig. 7

Temperatures in the UTOP accident

Grahic Jump Location
Fig. 4

Temperatures in the UOC accident

Grahic Jump Location
Fig. 5

Temperature reactivity feedback in the UOC accident

Grahic Jump Location
Fig. 6

Relative power and flow in the UTOP accident

Grahic Jump Location
Fig. 8

Temperature reactivity feedback in the UTOP accident

Grahic Jump Location
Fig. 9

Relative power and flow in the ULOF accident

Grahic Jump Location
Fig. 10

Temperatures in the ULOF accident

Grahic Jump Location
Fig. 11

Temperature reactivity feedback in the ULOF accident

Grahic Jump Location
Fig. 3

Relative power and mass flow rate in the UOC accident

Tables

Errata

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In