Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
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S. Qiu
, D. Zhang , L. Liu , M. Liu , R. Xu , C. Gong and G. H. Su
Abstract
Molten salt reactor (MSR) as one candidate of the Generation IV advanced nuclear power systems is attracted more attention in China due to its top ranked fuel cycle and thorium utilization. The MSRs are characterized by using liquid-fuel, which offers complicated coupling problem of neutronics and thermal hydraulics. In this paper, the fundamental model and numerical method are established to calculate and analyze the safety characteristics for liquid-fuel MSRs. The theories and methodologies are applied to the MOSART concept. The liquid-fuel flow effects on neutronics, reactivity coefficients and three operation parameters' influences at steady state are obtained, which provide the basic information for safety analysis. The unprotected loss of flow transient is calculated, the results of which shows the inherent safety characteristics of MOSART due to its strong negative reactivity feedbacks.
Kurzfassung
Wegen ihres Brennstoffkreislaufs und der Thorium-Nutzung sind Salzschmelze-Reaktoren in China von besonderem Interesse. Die Nutzung des flüssigen Brennstoffs führt zu komplizierteren gekoppelten Berechnungen der Neutronenkinetik und der Thermohydraulik im Vergleich zu Feststoffbrennstoff. In diesem Beitrag werden die Grundlagen zur Berechnung und Analyse der Sicherheit von Salzschmelze-Reaktoren mit flüssigem Brennstoff beschrieben. Diese basieren auf dem MOSART Konzept. Die im stationären Zustand bestimmten Kenndaten bilden dann die Grundlage für Sicherheitsanalysen. Am Beispiel der Berechnung einer Loss of Flow Transiente werden die inhärenten Sicherheitseigenschaften des MOSART Konzepts aufgrund seines streng negativen Reaktivitäts-Feedbacks dargestellt.
References
1 Bettis, E. S.; Schroeder, R. W.; Cristy, G. A.; et al.: The aircraft reactor experiment: design and construction. Nuclear Science and Engineering2 (1957) 804–82510.13182/NSE57-A35495Search in Google Scholar
2 Rosenthal, M. W.; Kasten, P. R.; Briggs, R. B.: Molten-salt reactors:history, states, and potential. Nuclear Application & Technology8 (1970) 107–11710.13182/NT70-A28619Search in Google Scholar
3 Forsberg, C. W., Renault, C., Brun, C. L.; et al.: Liquid Salt Applications and Molten Salt Reactors. Proc. of ICAPP’07, 13–18 May 2007, Nice, France; 200710.1051/rgn/20074063Search in Google Scholar
4 Ignatiev, V.; Feynberg, O.; Myasnikov, A.; et al.: Reactor Physics & Fuel Cycle Analysis of a Molten Salt Advanced Reactor Transmuter. ICAPP ’03, Cordoba, Spain, 4–7 May 2003Search in Google Scholar
5 Brovchenko, M.; Merle-Lucotte, E.; et al.: Molten Salt Fast Reactor transient analyses with the COUPLE code, ANS 2013 summer Meeting, Atlanta, USSearch in Google Scholar
6 Sahin, S.; Sahin, H. M.; Acir, A.: Criticality and burn up evolutions of the fixed bed nuclear reactor with alternative fuels. Energy Conversion and Management51 (2010) 1781–178710.1016/j.enconman.2009.12.044Search in Google Scholar
7 SINAP: Pre-conceptual Desigh of a 2 MW Pebble-bed Fluoride Salt Coolant High Temperature Test Reactor. Shanghai Insititute of Applied Phsics (SINAP), Chinese Academy of Sciences, 2012Search in Google Scholar
8 Forsberg, C. W.; Peterson, P. F.; PickardP. S.: Molten Salt Cooled Advanced High Temperature Reactor for Production of Hydrogen and Electricity. Nuclear Technology144 (2003) 289–30210.13182/nt03-1Search in Google Scholar
9 Raluca, O.; Scarlat; Peterson, P. F.: The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts. Nuclear Instruments and Methods in Physics Research A, 733 (2014) 57–6410.1016/j.nima.2013.05.094Search in Google Scholar
10 Maschek, W.; Stanculescu, A.; Arien, B. et al.: Report on International Results of the IAEA CRP on ‘Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste’. Energy Conversion and Management49 (2008) 1810–181910.1016/j.enconman.2007.08.018Search in Google Scholar
11 Heuer, D.; MerlE-Lucotte, E.; Allibert, M. et al.: Towards the thorium fuel cycle with molten salt fast reactors. Annals of Nuclear Energy64 (2014) 421–42910.1016/j.anucene.2013.08.002Search in Google Scholar
12 Wang, S.; Rineiski, A.; Zhang, D.: Molten Salt Fast Reactor analyses with SIMMER-III Code. ANS 2013 summer Meeting, Atlanta, USSearch in Google Scholar
13 Zhang, D.; Qiu, S.; Liu, C. et al.: Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt. Chinese Physics C32 (8) 624–62810.1088/1674-1137/32/8/007Search in Google Scholar
14 Zhang, D.; Qiu, S.; Su, G. H. et al.: Development of a steady state analysis code for a molten salt reactor. Annals of Nuclear Energy36 (2009) 590–60310.1016/j.anucene.2009.01.004Search in Google Scholar
15 Zhang, D.; Qiu, S.; Su, G. H. et al.: Analysis on the neutron kinetics for a molten salt reactor. Progress in Nuclear Energy51 (2009) 624–63610.1016/j.pnucene.2008.11.008Search in Google Scholar
16 Zhang, D.; Rineiski, A.; and Qiu, S.: Comparison of Modeling Options for Delayed Neutron Precursor Movement in a Molten Salt Reactor. Trans. Am. Nucl. Soc.100 (2009) 639Search in Google Scholar
17 Zhang, D.; Rineiski, A.; and Qiu, S.: Kinetics Models for Safety Studies of Fluid-Fuel Reactors. Trans. Am. Nucl. Soc.104 (2011)10.1016/j.pnucene.2015.01.011Search in Google Scholar
18 Zhang, D.; Zhai, Z.-G.; Chen, X.-N; Wang, S.; Rineiski, A.: COUPLE, A Coupled Neutronics and Thermal-Hydraulics Code for Transient Analyses of Molten Salt Reactors. ANS 2013 summer Meeting, Atlanta, USSearch in Google Scholar
19 Briant, R. C.; Weinberg, A. M.: Molten fluorides as power reactor fuels. Nuclear Science and Engineering2 (1957) 797–80310.13182/NSE57-A35494Search in Google Scholar
20 Henry, A. F.: The Application of Reactor Kinetics to the Analysis of Experiments. Nuclear Science and Engineering3 (1958) 52–7010.13182/NSE58-1Search in Google Scholar
21 Ott, K.O.; Neuhold, R. J.: Introductory Nuclear Reactor Dynamics. American Nuclear Society. 1985. 1Search in Google Scholar
22 Zhang, D.; Rineiski, A.; Wang, C. et al.: Development of a kinetic model for safety studies of liuqui-fuel reactors. Progress in Nuclear Energy81 (2015) 104–11210.1016/j.pnucene.2015.01.011Search in Google Scholar
23 MacPhee, J.: The kinetics of circulating reactors. Nuclear Science and Engineering4 (1958) 588–59710.13182/NSE4-588-597Search in Google Scholar
24 Zhang, D.; Qiu, S.; Su, G. H. et al.: Development of a safety analysis code for molten salt reactors. Nuclear Engineering and Design239 (2009) 1778–1785. 10.1016/j.nucengdes.2009.08.020Search in Google Scholar
25 Zhang, D.; Qiu, S.; Su, G. H. et al.: Evaluation of static thermophysical properties of the ternary molten salt system Li, Na, Be/F based on the modified Peng-Robinson equation. Journal of Power and Energy Systems2 (2008) 826–83310.1299/jpes.2.826Search in Google Scholar
26 Wang, S.; Rineiski, A.; Maschek, W.: Molten salt related extensions of the SIMMER-III code and its application for a burner reactor. Nuclear Engineering and Design236 (2006) 1580–158810.1016/j.nucengdes.2006.04.022Search in Google Scholar
27 Yamamoto, T.; Mitachi, K.; Ikeuchi, K. et al.: Transient characteristics of small molten salt reactor during blockage accident. Heat Transfer Asian Research35 (2006) 434–45010.1002/htj.20123Search in Google Scholar
28 IAEA: CRP on studies of advanced reactor technology options for effective incineration of radioactive wasteSearch in Google Scholar
© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
- A spray cooling technique for spent fuel assembly stored in pool
- KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
- 10.3139/124.110680
- Validation of the ATHLET-SC code by trans-critical transient data
- Qualification of CFD-models for multiphase flows
- The reactor dynamics code DYN3D
- Critical flow phenomena and modeling in advanced nuclear safety technology
- 10.3139/124.110682
- Safety and security aspects in design of digital safety I&C in nuclear power plants
- Thermohydraulic safety issues for liquid metal cooled systems
- Design and safety analysis of the helium cooled solid breeder blanket for CFETR
- Qualification of pebble fuel for HTGRs
- High temperature reactors for cogeneration applications
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Selected contributions from 1th Sino-German Symposium on Fundamentals of Advanced Nuclear Safety Technology
- Technical Contributions/Fachbeiträge
- Scientific codes developed and used at GRS – Nuclear simulation chain
- Challenges on innovations of newly-developed safety analysis codes
- Validation of system codes for plant application on selected experiments
- Progress of Experimental Research on Nuclear Safety in NPIC
- Severe accident research activities at Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
- THAI experimental programme for containment safety assessment under severe accident conditions
- A spray cooling technique for spent fuel assembly stored in pool
- KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors
- Coupled neutronics/thermal-hydraulics and safety characteristics of liquid-fueled Molten Salt Reactors
- 10.3139/124.110680
- Validation of the ATHLET-SC code by trans-critical transient data
- Qualification of CFD-models for multiphase flows
- The reactor dynamics code DYN3D
- Critical flow phenomena and modeling in advanced nuclear safety technology
- 10.3139/124.110682
- Safety and security aspects in design of digital safety I&C in nuclear power plants
- Thermohydraulic safety issues for liquid metal cooled systems
- Design and safety analysis of the helium cooled solid breeder blanket for CFETR
- Qualification of pebble fuel for HTGRs
- High temperature reactors for cogeneration applications