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Advancing the safety design of heat pipe cooled reactors: a case study of lead liquid bath and monolith stainless steel core

  • Taiwo Saheed Yinusa , Huaping Mei EMAIL logo , Chao Chen , Isaac Kwasi Baidoo , Size Chen , Shuyong Liu and Taosheng Li
Published/Copyright: April 9, 2025
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Abstract

Driven by the demand for compact, portable, clean, and environmentally friendly energy sources, heat pipe-cooled microreactors have gained increasing popularity and research interest. Over the past five decades, several design alternatives have been proposed. One of the most popular designs is the monolith stainless steel core structure. Others have sought to improve safety and efficiency by introducing liquid metal cores (Na/Pb) with promising superior safety and high energy output advantages. In this work, we advance the development of heat pipe-cooled reactors through a comparative case study of 300 kW monolith stainless steel (solid core) and a lead liquid bath (liquid core). A detailed comparative analysis was conducted to evaluate their neutronic and thermal performance. The Neutronic analysis was conducted using the SuperMC-CAS Monte Carlo code, while ANSYS (CFD) was used for thermal analysis. The results confirmed the feasibility of the stainless steel design and revealed the inherent advantages of the lead liquid core. Both cores exhibit sufficient excess reactivity to sustain reactor operation for 10 years without refueling. The lead liquid core enhances inherent safety, exhibits high negative reactivity feedback, and ensures a more uniform power distribution. Thermal analysis reveals that the lead liquid core maintains lower peak temperatures and gradients during normal operation and heat pipe failure, minimizing thermal stress and enhancing safety margins. This study affirmed the robustness of a proposed lead liquid core as a viable alternative to conventional designs providing enhanced safety and performance for heat pipe microreactors.


Corresponding author: Huaping Mei, Institute of Nuclear Energy Safety Technology, Hefei Institute of Physical Science, Chinese Academy of Sciences, Hefei, 230031, China, Email:

Funding source: HFIPS Director’s Fund

Award Identifier / Grant number: YZJJ202305-TS

Award Identifier / Grant number: YZJJ202406-TS

Funding source: Anhui Provincial Key Research and Development Project

Award Identifier / Grant number: 2022107020018

Acknowledgments

The authors wish to express their gratitude for the support for this work by the Anhui Provincial Key Research and Development Project (No. 2022107020018), and the HFIPS Director’s Fund (No. YZJJ202305-TS, No. YZJJ202406-TS). We acknowledge the joint support of the Chinese Academy of Sciences and World Academy of Science (CAS-TWAS).

  1. Research ethics: Not applicable.

  2. Informed consent: Not applicable.

  3. Author contributions: All authors have accepted responsibility for the entire content of this manuscript and approved its submission.

  4. Use of Large Language Models, AI and Machine Learning Tools: None declared.

  5. Conflict of interest: All other authors state no conflict of interest.

  6. Research funding: Anhui Provincial Key Research and Development Project (No. 2022107020018) and the HFIPS Director’s Fund (No. YZJJ202305-TS, No. YZJJ202406-TS).

  7. Data availability: Not applicable.

References

Ahmad, S., Chang, B., Li, B., Yang, Q., and Liu, C. (2021). Mass optimization of the radiation shadow shield for space nuclear power system. Prog. Nucl. Energy 131: 103607, https://doi.org/10.1016/j.pnucene.2020.103607.Search in Google Scholar

Alawneh, L.M., Vaghetto, R., Hassan, Y., and Harold, G. (2022). Conceptual design of a 3 MWth yttrium hydride moderated heat pipe cooled micro reactor. Nucl. Eng. Des. 397: 111931, https://doi.org/10.1016/j.nucengdes.2022.111931.Search in Google Scholar

Aldebie, F., Fernandez-Cosials, K., and Hassan, Y. (2023). Neutronic and thermal analysis of heat pipe cooled graphite moderated micro reactor. Nucl. Eng. Des. 413: 112483, https://doi.org/10.1016/j.nucengdes.2023.112483.Search in Google Scholar

Allen, T. and Crawford, D. (2007). Lead-cooled fast reactor systems and the fuels and materials challenges. Science and Technology of Nuclear Installations 2007: 097486, https://doi.org/10.1155/2007/97486.Search in Google Scholar

Baidoo, I., Li, B., Wu, B., Hao, L., Song, J., and Shitsi, E. (2022). Reactor pressure vessel damage (dpa/s) calculation and testing of 56Fe data libraries based on PCA benchmark model simulations using the SuperMC 3.4 code. Ann. Nucl. Energy 166: 108694, https://doi.org/10.1016/j.anucene.2021.108694.Search in Google Scholar

Bobkov, V. (2008). Thermophysical properties of materials for nuclear engineering: a tutorial and collection of data international. Atomic Energy Agency, Vienna.Search in Google Scholar

Brown, D.A., Chadwick, M.B., Capote, R., Kahler, A., Trkov, A., Herman, M., Sonzogni, A., Danon, Y., Carlson, A., Dunn, M., et al.. (2018). ENDF/B-VIII. 0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data. Nucl. Data Sheets 148: 1–142, https://doi.org/10.1016/j.nds.2018.02.001.Search in Google Scholar

Cai, B., Chen, S., Wu, T., Wang, Y., Zhang, Z., Yuan, C., Zhang, C., and Ge, Y. (2021). Neutronic design for heat pipe reactor with annular and accident tolerant fuels. Front. Energy Res. 9: 678262, https://doi.org/10.3389/fenrg.2021.678262.Search in Google Scholar

Cao, H. and Wang, G. (2019). The research on the heat transfer of a solid-core nuclear reactor cooled by heat pipe through a numerical simulation, considering the assembly gaps. Ann. Nucl. Energy 130: 431–439, https://doi.org/10.1016/j.anucene.2019.03.013.Search in Google Scholar

Chen, H., Qiao, S., Zhao, C., Zhang, T., and Yao, X. (2024). Design of moderator and collimator based on compact DT neutron source for neutron imaging. J. Nucl. Sci. Technol. 61: 1232–1238, https://doi.org/10.1080/00223131.2024.2310573.Search in Google Scholar

Feng, K., Hu, J., Wang, Y., Cong, T., Gu, H., and Guo, H. (2024). Multiphysics analysis of a metal hydride moderated megawatt heat pipe reactor with burnable poisons. Front. Energy Res. 12: 1346311, https://doi.org/10.3389/fenrg.2024.1346311.Search in Google Scholar

Feng, K., Wu, Y., Hu, J., Jin, X., Gu, H., and Guo, H. (2022). Preliminary analysis of a zirconium hydride moderated megawatt heat pipe reactor. Nucl. Eng. Des. 388: 111622, https://doi.org/10.1016/j.nucengdes.2021.111622.Search in Google Scholar

Guo, H., Feng, K., Gu, H., Yao, X., and Bo, L. (2021). Neutronic modeling of megawatt-class heat pipe reactors. Ann. Nucl. Energy 154: 108140, https://doi.org/10.1016/j.anucene.2021.108140.Search in Google Scholar

Lee, D.H. and Bang, I.C. (2024). Hybrid heat pipe shutdown rod as a novel concept of passive safety system for microreactor. Int. J. Energy Res. 2024: 6788210, https://doi.org/10.1155/2024/6788210.Search in Google Scholar

Lee, S.N., Tak, N.-I., and Choi, S.H. (2020). Gap assessment on the moderator temperature distribution in a space reactor. In: Transactions of the Korean Nuclear Society Spring Meeting. Korean Nuclear Society, pp. 21–22.Search in Google Scholar

Li, J., Cai, J., and Li, X. (2023). Conceptual design and feasibility analysis of a megawatt level low enriched uranium heat pipe cooled reactor core. Ann. Nucl. Energy 181: 109576, https://doi.org/10.1016/j.anucene.2022.109576.Search in Google Scholar

Li, S., Liang, Z., and Yan, B. (2022). A medium temperature heat pipe cooled reactor. Ann. Nucl. Energy 172: 109068, https://doi.org/10.1016/j.anucene.2022.109068.Search in Google Scholar

Ma, Y., Chen, E., Yu, H., Zhong, R., Deng, J., Chai, X., Huang, S., Ding, S., and Zhang, Z. (2020). Heat pipe failure accident analysis in megawatt heat pipe cooled reactor. Ann. Nucl. Energy 149: 107755, https://doi.org/10.1016/j.anucene.2020.107755.Search in Google Scholar

Ma, Y., Liu, J., Yu, H., Tian, C., Huang, S., Deng, J., Chai, X., Liu, Y., and He, X. (2022). Coupled irradiation-thermal-mechanical analysis of the solid-state core in a heat pipe cooled reactor. Nucl. Eng. Technol. 54: 2094–2106, https://doi.org/10.1016/j.net.2022.01.002.Search in Google Scholar

Ma, Y., Liu, M., Xie, B., Han, W., Yu, H., Huang, S., Chai, X., Liu, Y., and Zhang, Z. (2021a). Neutronic and thermal-mechanical coupling analyses in a solid-state reactor using Monte Carlo and finite element methods. Ann. Nucl. Energy 151: 107923, https://doi.org/10.1016/j.anucene.2020.107923.Search in Google Scholar

Ma, Y., Tian, C., Yu, H., Zhong, R., Zhang, Z., Huang, S., Deng, J., Chai, X., and Yang, Y. (2021b). Transient heat pipe failure accident analysis of a megawatt heat pipe cooled reactor. Prog. Nucl. Energy 140: 103904, https://doi.org/10.1016/j.pnucene.2021.103904.Search in Google Scholar

Mcclure, P.R., Poston, D.I., Dasari, V.R., and Reid, R.S. (2015). Design of megawatt power level heat pipe reactors. LA-UR-15-28840. Report of Los Alamos National Laboratory, Los Alamos, NM.10.2172/1226133Search in Google Scholar

Niederauer, G. and Lantz, E. (1970). A split-core heat-pipe reactor for space power applications. Report of NASA Lewis Research Center Cleveland, OH, United States. Report No.: NASA TM X-52918. https://www.osti.gov/biblio/7366283.Search in Google Scholar

Rivai, A.K. and Takahashi, M. (2008). Compatibility of surface-coated steels, refractory metals and ceramics to high temperature lead–bismuth eutectic. Prog. Nucl. Energy 50: 560–566, https://doi.org/10.1016/j.pnucene.2007.11.081.Search in Google Scholar

Rivai, A.K. and Takahashi, M. (2010). Corrosion investigations of Al–Fe-coated steels, high Cr steels, refractory metals and ceramics in lead alloys at 700° C. J. Nucl. Mat. 398: 146–152, https://doi.org/10.1016/j.jnucmat.2009.10.025.Search in Google Scholar

Sterbentz, J.W., Werner, J.E., Hummel, A.J., Kennedy, J.C., O’Brien, R.C., Dion, A.M., Wright, R.N., and Ananth, K.P. (2017). Preliminary assessment of two alternative core design concepts for the special purpose reactor, (No.INL/EXT-17-43212). Idaho National Lab (INL), Idaho Falls, ID.10.2172/1413987Search in Google Scholar

Wang, D., Yan, B., and Chen, J. (2020). The opportunities and challenges of micro heat piped cooled reactor system with high efficiency energy conversion units. Ann. Nucl. Energy 149: 107808, https://doi.org/10.1016/j.anucene.2020.107808.Search in Google Scholar

Wu, A., Wang, W., Zhang, K., Shen, S., Duan, W., Pan, R., Luo, X., and Chen, H. (2023a). Multiphysics coupling analysis of heat pipe reactor based on OpenMC and COMSOL Multiphysics. Ann. Nucl. Energy 194: 110115, https://doi.org/10.1016/j.anucene.2023.110115.Search in Google Scholar

Wu, Y., Zheng, Y., Chen, Q., Li, J., Du, X., Wang, Y., and Tao, Y. (2024). Conceptual design of a MW heat pipe reactor. Nucl.Eng. Technol. 56: 1116–1123, https://doi.org/10.1016/j.net.2024.02.009.Search in Google Scholar

Wu, Y., Zheng, Y., Tao, Y., Liu, X., Du, X., and Wang, Y. (2023b). The low-enriched uranium core design of a MW heat pipe cooled reactor. Nucl. Eng. Des. 404: 112195, https://doi.org/10.1016/j.nucengdes.2023.112195.Search in Google Scholar

Xiao, Z., Liu, J., Jiang, Z., and Luo, L. (2022). Corrosion behavior of refractory metals in liquid lead at 1000 °C for 1000 h. Nucl.Eng. Technol. 54: 1954–1961, https://doi.org/10.1016/j.net.2021.12.014.Search in Google Scholar

Yan, B., Wang, C., and Li, L. (2020). The technology of micro heat pipe cooled reactor: a review. Ann. Nucl. Energy 135: 106948, https://doi.org/10.1016/j.anucene.2019.106948.Search in Google Scholar

Yinusa, T.S., Mei, H.-P., Chen, C., Baidoo, I.K., Liu, S., and Li, T. (2024). Core design and neutronics analysis of lead liquid metal-based heat pipe cooled reactor for undersea purpose. Kerntechnik 90: 231–238, https://doi.org/10.1515/kern-2024-0110.Search in Google Scholar

Yu, L., Wang, L., Liu, C., Wu, B., Song, J., and Wu, Y. (2020). Development and testing of a coupled SuperMC and SUBCHANFLOW code for LWR simulations. Ann. Nucl. Energy 144: 107465, https://doi.org/10.1016/j.anucene.2020.107465.Search in Google Scholar

Zare Ganjaroodi, S., Khameh, H., Kazem Farahzadi, M., Kasesaz, Y., and Zarifi, E. (2023). Neutronic and safety analysis of CAREM-25 small modular reactor using supermc advanced code. Int. J. Reliabil. Risk Saf. Theory Appl. 6: 107–113, https://doi.org/10.22034/IJRRS.2023.6.2.12.Search in Google Scholar

Zhuang, D. (2023). Criticality safety calculation and analysis for NPP transportation of fuel assemblies. In: Proceedings of the 1st international conference on data processing, control and simulation, 1. Science and Technology Publications, Southampton, UK, pp. 51–57.10.5220/0012145900003562Search in Google Scholar

Received: 2024-12-27
Accepted: 2025-03-10
Published Online: 2025-04-09
Published in Print: 2025-06-26

© 2025 Walter de Gruyter GmbH, Berlin/Boston

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