Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5

Cheng-Long Wang; Tian-Cai Liu; Si-Miao Tang; Wen-Xi Tian; Sui-Zheng Qiu; Guang-Hui Su

doi:10.1007/s41365-018-0537-3

Your Location：

Home >

Browse articles >

Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5

NUCLEAR ENERGY SCIENCE AND ENGINEERING | Updated：2021-02-01

- Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5
- Nuclear Science and Techniques Vol. 30, Issue 1, Article number: 12(2019)
- Affiliations：
  
  1.Department of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, China
  2.China Institute of Atomic Energy, Beijing 102413, China
- Author bio：
  
  Corresponding author: szqiu@mail.xjtu.edu.cn
- Funds：
- DOI：10.1007/s41365-018-0537-3
  CLC：
Scan for full text
Cheng-Long Wang, Tian-Cai Liu, Si-Miao Tang, et al. Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5. [J]. Nuclear Science and Techniques 30(1):12(2019)
DOI：

Cheng-Long Wang, Tian-Cai Liu, Si-Miao Tang, et al. Thermal-hydraulic analysis of space nuclear reactor TOPAZ-II with modified RELAP5. [J]. Nuclear Science and Techniques 30(1):12(2019) DOI： 10.1007/s41365-018-0537-3.

摘要

Abstract

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5 (RELAP5), with the implementation of sodium-potassium eutectic alloy (NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal-hydraulic characteristics of the space nuclear reactor TOPAZ-II. A RELAP5 model including thermionic fuel elements (TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector, moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-II system model, an electrical ground test of the fully integrated TOPAZ-II system, the V-71 unit, is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8% and 10%, respectively. The detailed thermal-hydraulic characteristics of TOPAZ-II are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

关键词

Keywords

Space nuclear reactor TOPAZ-IIThermal-hydraulic analysisRELAP5 modification

references

L.S. Mason, A comparison of Brayton and Stirling space nuclear power systems for power levels from 1 kilowatt to 10 megawatts. AIP Conference Proceedings: Space Technology and Applications International Forum (STAIF-2000) on Space Exploration and Transportation - Journey into the Future. 552(1). 1017-1022 (2001). DOI: 10.1063/1.1358045http://doi.org/10.1063/1.1358045

G.L. Bennett, R.J. Hemler, A. Schock, Space nuclear power: An overview. J. Propul. Power. 12(5), 901-910 (1996). DOI: 10.2514/3.24121http://doi.org/10.2514/3.24121

D. Buden, Summary of space nuclear reactor power systems (1983-1992). Idaho National Engineering Laboratory Report (1993), Report ID: INEL/MISC-93085. https://inis.iaea.org/collection/NCLCollectionStore/_Public/25/070/25070118.pdfhttps://inis.iaea.org/collection/NCLCollectionStore/_Public/25/070/25070118.pdf

A.J. Glassman, Summary of Brayton cycle analytical studies for space-power system applications. Sol. Energy. 7 (1964). https://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19640020187.pdfhttps://ntrs.nasa.gov/archive/nasa/casi.ntrs.nasa.gov/19640020187.pdf

R.S Dabrowski, US-Russian Cooperation in Science and Technology: A Case Study of the TOPAZ Space-Based Nuclear Reactor International Program. Connections the Quarterly Journal. 13(1), 71-87 (2013). http://procon.bg/bg/system/files/13.1.05_dabrowski.pdfhttp://procon.bg/bg/system/files/13.1.05_dabrowski.pdf

S.S. Voss, E.A. Rodriguez, Russian TOPAZ II System Test Program (1970–1989). AIP Conference Proceedings: 11th Symposium on Space Nuclear Power and Propulsion. 301(1), 803-812 (1994). DOI: 10.1063/1.2950273http://doi.org/10.1063/1.2950273

J.F. Fairchild, J.P. Koonmen, F.V. Thome, Thermionic system evaluated test (TSET) facility description. AIP Conference Proceedings: Ninth Symposium on Space Nuclear Power Systems. 246(1), 836-842 (1992). DOI: 10.1063/1.41884http://doi.org/10.1063/1.41884

V.P. Nicitin, B.G. Ogloblin, A.N. Luppov et al., Special features and results of the "TOPAZ II" nuclear power system tests with electric heating. AIP Conference Proceedings: Ninth Symposium on Space Nuclear Power Systems. 246(1), 41-46 (1992). DOI: 10.1063/1.41803http://doi.org/10.1063/1.41803

M.S. El‐Genk, H. Xue, C. Murray et al., TITAM thermionic integrated transient analysis model: Load‐following of a single‐cell TFE. AIP Conference Proceedings: Ninth Symposium on Space Nuclear Power Systems. 246(1), 1013-1022 (1992). DOI: 10.1063/1.41776http://doi.org/10.1063/1.41776

M.S. El‐Genk, D.V. Paramonov, A.C. Marshall, Startup Simulation of the TOPAZ‐II Reactor System for Accident Conditions. AIP Conference Proceedings: 11th Symposium on Space Nuclear Power and Propulsion. 301(1), 1059-1068 (1994). DOI: 10.1063/1.2950104http://doi.org/10.1063/1.2950104

M.S. El‐Genk, H. Xue, D.V. Paramonov. Start‐up simulation of a thermionic space nuclear reactor system. AIP Conference Proceedings: 10th Symposium on Space Nuclear Power and Propulsion. 271(2), 935-950 (1993). DOI: 10.1063/1.43119http://doi.org/10.1063/1.43119

S.S. Voss, The TOPAZ II space reactor response under accident conditions. Presented at the International Topical Meeting on the Safety of Advanced Reactors, Pittsburgh, PA, 18-20 Apr. 1994. https://www.osti.gov/biblio/10115301https://www.osti.gov/biblio/10115301

V.H. Standley, D.B. Morris, M.J. Schuller, CENTAR modeling of the TOPAZ‐II: Loss of vacuum chamber cooling during full power ground test. AIP Conference Proceedings: Ninth Symposium on Space Nuclear Power Systems. 246(1), 1129-1134 (1992). DOI: 10.1063/1.41770http://doi.org/10.1063/1.41770

W.W. Zhang, Z.Y. Ma, D.L. Zhang, et al., Transient thermal–hydraulic analysis of a space thermionic reactor. Ann. Nucl. Energy. 89, 38-49 (2016). DOI: 10.1016/j.anucene.2015.10.035http://doi.org/10.1016/j.anucene.2015.10.035

W.W. Zhang, D.L. Zhang, W.X. Tian, et al. Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator. Nucl. Eng. Des. 307, 218-233 (2016). DOI: 10.1016/j.nucengdes.2016.07.020http://doi.org/10.1016/j.nucengdes.2016.07.020

S.S. Voss, TOPAZ-II System Description. Proceedings of the Fourth International Conference on Engineering, Construction, and Operations in Space, Albuquerque, NM. 717-728(1994). https://inis.iaea.org/collection/NCLCollectionStore/_Public/25/040/25040534.pdfhttps://inis.iaea.org/collection/NCLCollectionStore/_Public/25/040/25040534.pdf

C.B. Jackson, Liquid Metals Handbook. Sodium-NaK Supplement. Atomic Energy Commission, Washington, DC;

Bureau of Ships, Report ID: TID-5277 (1955). https://www.osti.gov/biblio/4376530https://www.osti.gov/biblio/4376530

Z.Y. Qian, Thermal Properties of Low Melting Point Metals. (Science Press, China, 1985), In Chinese

O.J. Foust, Sodium-NaK engineering handbook. (Gordon & Breach Science Pub., 1972), p.102.

M. Zemansky, R. Dittman, Heat and Thermodynamics. (McGraww-Hill Inc. Pub., 1997). p. 53

S. Aoki, O.E. Dwyer, Current liquid-metal heat transfer research in Japan. Tokyo Inst. of Tech., Report ID: NSA-29-000316. 1973. https://www.osti.gov/biblio/4423836https://www.osti.gov/biblio/4423836

O.E. Dwyer, P.S. Tu, Bilateral heat transfer to liquid metals flowing turbulently through annuli. Nucl. Sci. Eng., 21(1), 90-105 (1965). DOI: 10.13182/NSE65-A21018http://doi.org/10.13182/NSE65-A21018

N.E. Todreas, M.S. Kazimi, Nuclear systems: thermal hydraulic fundamentals. (CRC press, 2012), p.306.

D.V. Paramonov, M.S. El‐Genk, Comparison of a TOPAZ‐II Model with Experimental Data from the V‐71 Unit Tests. AIP Conference Proceedings: 11th Symposium on Space Nuclear Power and Propulsion. 301(1), 829-843 (1994). DOI: 10.1063/1.2950277http://doi.org/10.1063/1.2950277

Views

Downloads

CSCD

Alert me when the article has been cited

Submit

Tools

Publicity Resources

Thermal hydraulic characteristics of helical coil once-through steam generator under ocean conditions

Related Author

No data

Related Institution

School of Nuclear Science and Technology, University of Science and Technology of China, No. 96, Jinzhai Road, Baohe District

Address：Editorial Office of NST, 2019 Jialuo Road, Jiading, Shanghai 201800, China Postal code：201800
Email：nst@sinap.ac.cn
Technical support is provided by Beijing Founder electronics co., LTD 沪ICP备05005479号-1 京公网安备11010802024621
It is recommended to read the content of this site in Chrome&IE9+. Please switch to extreme mode in browser 360.
Cookies We use cookies to help provide and enhance our service and tailor content. By continuing, you agree to the use of cookies.

⁰