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Temperature analysis of the HTR-10 after full load rejection

  • Xiang-Cheng Zhang
  • Ding She
  • Fu-Bing ChenEmail author
  • Lei Shi
  • Yu-Jie Dong
  • Zuo-Yi Zhang
Article
  • 17 Downloads

Abstract

The HTR-10 is a small modular high-temperature reactor located in Tsinghua University, in China. After the reactor ran continuously for 72 h at full power in the commissioning stage, a full load rejection test was conducted by manually disconnecting the generator from the grid. In this study, reactor transients were analyzed using the THERMIX model. Some of the important thermal–fluid phenomena that occurred after the test initiation are discussed here, including the natural convection of helium in the core and the temperature redistribution in the reactor. Temperatures reproduced for the measuring points arranged in the internals were in agreement with the test data. This demonstrated that the code and calculation model are suitable for posttest analysis applications. Regarding the safety features of the reactor, there was a large margin between the predicted maximum fuel temperature of 997 °C and the safety limit of 1620 °C.

Keywords

HTR-10 Load rejection THERMIX Code validation 

References

  1. 1.
    J. Bickel, Grid stability and safety issues associated with nuclear power plants, Paper Presented at the Workshop on Power Grid Interconnection in Northeast Asia, Beijing, May 14–16, 2001, China. http://www.nautilus.org/wp-content/uploads/2012/01/Bickel.pdf
  2. 2.
    B. Zerger, M. Noël, Nuclear power plant commissioning experience. Prog. Nucl. Energy 53, 668–672 (2011).  https://doi.org/10.1016/j.pnucene.2011.04.010 CrossRefGoogle Scholar
  3. 3.
    W. Sun, S. Zhou, A design of in-plant electric power system for the HTR-10. Nucl. Eng. Des. 218, 227–232 (2002).  https://doi.org/10.1016/S0029-5493(02)00194-2 CrossRefGoogle Scholar
  4. 4.
    Z. Gao, L. Shi, Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core. Nucl. Eng. Des. 218, 51–64 (2002).  https://doi.org/10.1016/S0029-5493(02)00198-X CrossRefGoogle Scholar
  5. 5.
    Z. Gao, L. Shi, Thermal hydraulic transient analysis of the HTR-10. Nucl. Eng. Des. 218, 65–80 (2002).  https://doi.org/10.1016/S0029-5493(02)00199-1 CrossRefGoogle Scholar
  6. 6.
    J. Wolters, G. Breitbach, R. Moormann, Air and water ingress accidents in a HTR-Modul of side-by-side concept, in Gas-cooled Reactor Safety and Accident Analysis (Proceedings of a Specialists’ Meeting, Oak Ridge, May 1315, 1985, USA). (IAEA-TECDOC-358, Vienna, 1985). https://www-pub.iaea.org/books/iaeabooks/613/Gas-cooled-Reactor-Safety-and-Accident-Analysis-Proceedings-of-a-Specialists-Meeting-Oak-Ridge-13-15-May-1985
  7. 7.
    R. Hedrick, J. Cleveland, BLAST: a digital computer program for the dynamic simulation for the high temperature gas cooled reactor reheater-steam generator module. (ORNL/NUREG/TM-38, Oak Ridge National Laboratory, TN, USA, 1976). https://www.researchgate.net/publication/236550781_BLAST_a_digital_computer_program_for_the_dynamic_simulation_of_the_high_temperature_gas_cooled_reactor_reheater-steam_generator_module
  8. 8.
    J. Cleveland, S. Greene, Application of THERMIX-KONVEK code to accident analyses of modular pebble bed high temperature rectors (HTRs) (ORNL/TM-9905, Oak Ridge National Laboratory, USA, 1986). https://www.osti.gov/scitech/biblio/5150133
  9. 9.
    K. Kruger, A. Bergerfurth, S. Burger et al., Preparation, conduct, and experimental results of the AVR loss-of-coolant accident simulation test. Nucl. Sci. Eng. 107, 99–113 (1991).  https://doi.org/10.13182/nse91-a15725 CrossRefGoogle Scholar
  10. 10.
    L. Wolf, W. Scherer, W. Giesser et al., High temperature reactor core physics and reactor dynamics. Nucl. Eng. Des. 121, 227–240 (1990).  https://doi.org/10.1016/0029-5493(90)90108-A CrossRefGoogle Scholar
  11. 11.
    H. Haque, W. Feltes, G. Brinkmann, Thermal response of a modular high temperature reactor during passive cooldown under pressurized and depressurized conditions. Nucl. Eng. Des. 236, 475–484 (2006).  https://doi.org/10.1016/j.nucengdes.2005.10.027 CrossRefGoogle Scholar
  12. 12.
    International Atomic Energy Agency, Heat transport and afterheat removal for gas cooled reactors under accident conditions (TECDOC-1163, Vienna, 2000). https://www-pub.iaea.org/MTCD/Publications/PDF/te_1163_prn.pdf
  13. 13.
    International Atomic Energy Agency, Evaluation of high temperature gas cooled reactor performance: Benchmark analysis related to initial testing of the HTTR and HTR-10 (TECDOC-1382, Vienna, 2003). https://www-pub.iaea.org/MTCD/Publications/PDF/te_1382_web/attention.pdf
  14. 14.
    International Atomic Energy Agency, Evaluation of high temperature gas cooled reactor performance: Benchmark analysis related to the PBMR-400, PBMM, GT-MHR, HTR-10 and the ASTRA critical facility (TECDOC-1694, Vienna, 2013). https://www-pub.iaea.org/MTCD/Publications/PDF/TE-1694_web.pdf
  15. 15.
    F. Chen, Y. Dong, Y. Zheng et al., Benchmark calculation for the steady-state temperature distribution of the HTR-10 under full-power operation. J. Nucl. Sci. Technol. 46, 572–580 (2009).  https://doi.org/10.1080/18811248.2007.9711564 CrossRefGoogle Scholar
  16. 16.
    F. Chen, Y. Dong, Z. Zhang et al., Post-test analysis of helium circulator trip without scram at 3 MW power level on the HTR-10. Nucl. Eng. Des. 239, 1010–1018 (2009).  https://doi.org/10.1016/j.nucengdes.2009.02.009 CrossRefGoogle Scholar
  17. 17.
    F. Chen, Y. Dong, Z. Zhang, Temperature response of the HTR-10 during the power ascension test. Sci. Technol. Nucl. Instal. 2015, 1–13 (2015).  https://doi.org/10.1155/2015/302648 CrossRefGoogle Scholar
  18. 18.
    F. Chen, Y. Dong, Z. Zhang, Post-test simulation of the HTR-10 reactivity insertion without scram. Ann. Nucl. Energy 2016, 36–45 (2016).  https://doi.org/10.1016/j.anucene.2016.01.023 CrossRefGoogle Scholar
  19. 19.
    Z. Wu, D. Lin, D. Zhong, The design features of the HTR-10. Nucl. Eng. Des. 218, 25–32 (2002).  https://doi.org/10.1016/S0029-5493(02)00182-6 CrossRefGoogle Scholar
  20. 20.
    S. Zhong, S. Hu, M. Zha et al., Thermal hydraulic instrumentation system of the HTR-10. Nucl. Eng. Des. 218, 199–208 (2002).  https://doi.org/10.1016/S0029-5493(02)00191-7 CrossRefGoogle Scholar
  21. 21.
    M. Zha, S. Zhong, R. Chen et al., Temperature measuring system of the in-core components for Chinese 10 MW high temperature gas-cooled reactor. J. Nucl. Sci. Technol. 39, 1086–1093 (2002).  https://doi.org/10.1080/18811248.2002.9715297 CrossRefGoogle Scholar
  22. 22.
    R. Li, H. Ju, Structural design and two-phase flow stability test for the steam generator. Nucl. Eng. Des. 218, 179–187 (2002).  https://doi.org/10.1016/S0029-5493(02)00189-9 CrossRefGoogle Scholar
  23. 23.
    S.Y. Hu, X.H. Liang, L.Q. Wei, Commissioning and operation experience and safety experiments on HTR-10, Paper Presented at the 3rd International Topical Meeting on High Temperature Reactor Technology (HTR2006), Johannesburg, October 1–4, 2006, South Africa Google Scholar
  24. 24.
    H. Reutler, G. Lohnert, Advantages of going modular in HTRs. Nucl. Eng. Des. 78, 129–136 (1984).  https://doi.org/10.1016/0029-5493(84)90298-x CrossRefGoogle Scholar
  25. 25.
    Z.Y. Zhang, Y.J. Dong, F. Li et al., The Shandong Shidao Bay 200 MWe high-temperature gas-cooled reactor pebble-bed module (HTR-PM) demonstration power plant: an engineering and technological innovation. Engineering 2, 112–118 (2016).  https://doi.org/10.1016/j.eng.2016.01.020 CrossRefGoogle Scholar

Copyright information

© China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, Chinese Nuclear Society and Springer Nature Singapore Pte Ltd. 2019

Authors and Affiliations

  • Xiang-Cheng Zhang
    • 1
  • Ding She
    • 1
  • Fu-Bing Chen
    • 1
    Email author
  • Lei Shi
    • 1
  • Yu-Jie Dong
    • 1
  • Zuo-Yi Zhang
    • 1
  1. 1.Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of EducationTsinghua UniversityBeijingChina

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