Core temperature distributions in HTR-10 operating at very high temperatures

SUN Shiyan; ZHANG Youjie; ZHENG Yanhua; XIA Bing

doi:10.16511/j.cnki.qhdxxb.2020.22.036

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Journal of Tsinghua University(Science and Technology) ›› 2021, Vol. 61 ›› Issue (11) : 1301-1307. DOI: 10.16511/j.cnki.qhdxxb.2020.22.036

NUCLEAR ENERGY AND NEW ENERGY

Core temperature distributions in HTR-10 operating at very high temperatures

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Abstract

The 10 MW high temperature gas-cooled reactor test module (HTR-10) has a relatively large fuel temperature margin when operating at full power and steady state for the rated condition; thus, the coolant outlet temperature can be further increased above 700℃, which is useful for pebble-bed very high temperature gas-cooled reactors (HTGR) to operate with high coolant temperature at the core outlet. Several steady-state very high temperature operating conditions were designed for HTR-10 in this study based on a heat balance between the core heating power and the heat carried by the coolant. An improved thermal hydraulic model was used to analyze the core temperature distribution at each condition for the initial core to determine the temperature distribution characteristics, and the maximum core temperature for the nonuniform fuel distribution was discussed. For the coolant outlet temperature of 1 000℃, the predicted maximum core temperature is still below the fuel temperature limit, even when the most serious nonuniform distribution of the fuel and graphite balls occurs.

Key words

pebble-bed high temperature gas-cooled reactor (HTGR) / 10 MW high temperature gas-cooled reactor test module (HTR-10) / operating at very high temperatures / thermal hydraulic analysis / core temperature distribution

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SUN Shiyan, ZHANG Youjie, ZHENG Yanhua, XIA Bing. Core temperature distributions in HTR-10 operating at very high temperatures[J]. Journal of Tsinghua University(Science and Technology). 2021, 61(11): 1301-1307 https://doi.org/10.16511/j.cnki.qhdxxb.2020.22.036

References

[1] 王捷. 高温气冷堆技术背景和发展潜力的初步研究[J]. 核科学与工程, 2002, 22(4):325-330.WANG J. Preliminary study on technical base and future potential of high temperature gas-cooled reactor[J]. Chinese Journal of Nuclear Science and Engineering, 2002, 22(4):325-330. (in Chinese)
[2] SCHULTEN R. The AVR nuclear power plant:A milestone in high-temperature reactor development[J]. Nuclear Engineering and Design, 1985, 90(4):388-390.
[3] BAUST E, RAUTENBERG J, WOHLER J. Results and experience from the commissioning of the THTR 300[J]. Atomkernenergie-Kerntechnik, 1985, 47:141-144.
[4] WU Z X, LIN D C, ZHONG D X. The design features of the HTR-10[J]. Nuclear Engineering and Design, 2002, 218(1-3):25-32.
[5] US DOE Nuclear Energy Research Advisory Committee, Generation IV International Forum. A technology roadmap for generation IV nuclear energy systems[R]. Washington DC, USA:US DOE Nuclear Energy Research Advisory Committee, Generation IV International Forum, 2002.
[6] 陈福冰. 利用HTR-10试验数据对安全分析程序THERMIX的验证[D]. 北京:清华大学, 2009.CHEN F B. Validation of the safety analysis code THERMIX by using the HTR-10 experimental data[D]. Beijing:Tsinghua University, 2009. (in Chinese)
[7] 徐小琳, 曲荣红, 边晖. 10 MW高温气冷实验堆首次装料和趋近临界[J]. 高技术通讯, 2001(3):104-106.XU X L, QU R H, BIAN H. The first loading and critical experiment for 10 MW high temperature gas-cooled reactor[J]. High Technology Letters, 2001(3):104-106. (in Chinese)
[8] 刘俊杰, 王敏稚, 张征明, 等. 10 MW高温气冷实验堆的堆体结构特点[J]. 核动力工程, 2001(1):53-56.LIU J J, WANG M Z, ZHANG Z M, et al. Features of reactor structure design for 10 MW high temperature gas-cooled reactor[J]. Nuclear Power Engineering, 2001(1):53-56. (in Chinese)
[9] GAO Z Y, SHI L. Thermal hydraulic calculation of the HTR-10 for the initial and equilibrium core[J]. Nuclear Engineering and Design, 2002, 218(1):51-64.
[10] CLEVELAND J C, GREENE S R. Application of THERMIX-KONVEK code to accident analyses of modular pebble bed high temperature reactors (HTRs)[R]. Oak Ridge, USA:Oak Ridge National Laboratory, 1986.
[11] SUN S Y, ZHANG Y J, ZHENG Y H. Research on influence of different simulation methods of bypass flow in thermal hydraulic analysis on temperature distribution in HTR-10[J]. Science and Technology of Nuclear Installations, 2020, 2020:4754589.
[12] KADAK A C, BAZANT M Z. Pebble flow experiments for pebble bed reactors[C]//Proceedings of 2^nd International Topical Meeting on High Temperature Reactor Technology. Beijing, China, 2004:H05.