Influence of <sup>235</sup>U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

Xiao-Xiao Li; De-Yang Cui; Yu-Wen Ma; Xiang-Zhou Cai; Jin-Gen Chen

doi:10.1007/s41365-019-0694-z

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Influence of ²³⁵U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

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

- Influence of ²³⁵U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor
- Nuclear Science and Techniques Vol. 30, Issue 11, Article number: 166(2019)
- Affiliations：
  
  1.Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China
  2.CAS Innovative Academies in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800, China
  3.University of Chinese Academy of Sciences, Beijing 100049, China
- Author bio：
  
  caixz@sinap.ac.cn
  chenjg@sinap.ac.cn
- Funds：
- DOI：10.1007/s41365-019-0694-z
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Xiao-Xiao Li, De-Yang Cui, Yu-Wen Ma, et al. Influence of ²³⁵U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor. [J]. Nuclear Science and Techniques 30(11):166(2019)
DOI：

Xiao-Xiao Li, De-Yang Cui, Yu-Wen Ma, et al. Influence of ²³⁵U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor. [J]. Nuclear Science and Techniques 30(11):166(2019) DOI： 10.1007/s41365-019-0694-z.

摘要

Abstract

To optimize the temperature coefficient of reactivity (TCR) for a graphite-moderated and liquid-fueled molten salt reactor, the effects of fuel salt composition on the fuel salt temperature coefficient of reactivity (FSTC) were investigated in our earlier work. In this study, we aim to provide a more comprehensive analysis of the TCR by considering the effects of the graphite moderator temperature coefficient of reactivity (MTC). The effects of ,235,U enrichment and heavy metal (HM) proportion in the salt mixture on the MTC are investigated from the perspective of the six-factor formula based on a full-core model. For the MTC (labeled " ,α,TM,"), the temperature coefficient of the fast fission factors (,α,TM,(,ε,)) is positive, while those of the resonance escape probability (,α,TM,(,p,)), the thermal reproduction factor (,α,TM,(,η,)), the thermal utilization factor (,α,TM,(,f,)), and the total non-leakage probability (,α,TM,(,Λ,)) are negative. The results reveal that the magnitudes of ,α,TM,(,ε,) and ,α,TM,(,p,) for the MTC are similar. Thus, variations in the MTC with ,235,U enrichment for different HM proportions are mainly dependent on ,α,TM,(,η,),α,TM,(,Λ,), and ,α,TM,(,f,), but especially on the former two. To obtain a more negative MTC, a lower HM proportion and/or a lower ,235,U enrichment is recommended. Together with our previous studies on the FSTC, a relatively soft neutron spectrum could strengthen the TCR with a sufficiently negative MTC.

关键词

Keywords

Molten salt reactor (MSR)Moderator temperature coefficient of reactivity (MTC)Six-factor formula

references

J. Serp, M. Allibert, O. Beneš et al., The molten salt reactor (MSR) in generation IV: overview and perspectives. Prog. Nucl. Energy. 77, 308 (2014). doi: 10.1016/j.pnucene.2014.02.014http://doi.org/10.1016/j.pnucene.2014.02.014

R. C. Robertson, MSRE design and operations report. Part I. Description of reactor design. Oak Ridge National Lab., ORNL-TM-0728, (1965). doi: 10.2172/4654707http://doi.org/10.2172/4654707

R. C. Robertson, Conceptual design study of a single-fluid molten-salt breeder reactor. Oak Ridge National Lab., ORNL-4541, (1971). doi: 10.2172/4030941http://doi.org/10.2172/4030941

D. K. L. Tsang, B. J. Marsden, S. L. Fok et al., Graphite thermal expansion relationship for different temperature ranges. Nucl. Technol. 43, 2902 (2005). doi: 10.1016/j.carbon.2005.06.009http://doi.org/10.1016/j.carbon.2005.06.009

J. Žáková, Analysis of an Advanced Graphite Moderated and Molten Salt Cooled High Temperature Reactor. Master of science thesis, Department of Reactor Physics, Royal Institute of Technology Stockholm, Sweden (2006).

D. Y. Cui, S. P. Xia, X. X. Li et al., Transition toward thorium fuel cycle in a molten salt reactor by using plutonium. Nucl. Sci. Tech. 28, 152 (2017). doi: 10.1007/s41365-017-0303-yhttp://doi.org/10.1007/s41365-017-0303-y

J. Křepel, U. Rohde, U. Grundmann et al., Dynamics of molten salt reactors. Nucl. Technol. 164, 34 (2008). doi: 10.13182/NT08-A4006http://doi.org/10.13182/NT08-A4006

L. Mathieua, D. Heuera, R. Brissota et al., The thorium molten salt reactor: Moving on from the MSBR. Prog. Nucl. Energy. 48, 664 (2006). doi: 10.1016/j.pnucene.2006.07.005http://doi.org/10.1016/j.pnucene.2006.07.005

L. Mathieu, D. Heuer, E. Merle-Lucotte et al., Possible configurations for the thorium molten salt reactor and advantages of the fast nonmoderated version. Nucl. Sci. Eng. 161, 78 (2009). doi: 10.13182/NSE07-49http://doi.org/10.13182/NSE07-49

K. Nagy, J.L. Kloosterman, D. Lathouwers et al., New breeding gain definitions and their application to the optimization of a molten salt reactor design. Ann. Nucl. Energy 38, 601(2011). doi: 10.1016/j.anucene.2010.09.024http://doi.org/10.1016/j.anucene.2010.09.024

K. Nagy, Dynamics and fuel cycle analysis of a moderated molten salt reactor. Delft University of Technology (2012). doi: 10.4233/uuid:b4d5089d-c2de-446b-94cf-c563dd73e8f1http://doi.org/10.4233/uuid:b4d5089d-c2de-446b-94cf-c563dd73e8f1

J. Křepel, B. Hombourger, C. Fiorina et al., Fuel cycle advantages and dynamics features of liquid fueled MSR. Ann. Nucl. Energy 64, 380 (2014). doi: 10.1016/j.anucene.2013.08.007http://doi.org/10.1016/j.anucene.2013.08.007

C. Y. Zou, X. Z. Cai, D. Z. Jiang, et al., Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor. Nucl. Eng. Des. 281, 114 (2015). doi: 10.1016/j.nucengdes.2014.11.022http://doi.org/10.1016/j.nucengdes.2014.11.022

F. Lantelme, H. Groult, Molten salts chemistry: from lab to applications. Elsevier (2013). doi: 10.1016/B978-0-12-398538-5.09995-9http://doi.org/10.1016/B978-0-12-398538-5.09995-9

D. F. Williams, K. T. Clarno, Evaluation of Salt Coolants for Reactor Applications, Nucl. Technol. 163:3, 330 (2008). doi: 10.13182/NT08-A3992http://doi.org/10.13182/NT08-A3992

A. Nuttin, D. Heuer, A. Billebaud et al., Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production, Prog. Nucl. Energy. 46, 77 (2005). doi: 10.1016/j.pnucene.2004.11.001http://doi.org/10.1016/j.pnucene.2004.11.001

P. N. Haubenreich, J. R. Engel, B. E. Engel, et al., MSRE design and operations report. Par III. Nuclear analysis. Oak Ridge National Lab., ORNL-TM-0730, (1964).doi: 10.2172/4114686http://doi.org/10.2172/4114686

E. Merle-Lucotte, D. Heuer, M. Allibert, et al., Optimized transition from the reactors of second and third generations to the Thorium Molten Salt Reactor. ICAPP 2007: International Congress on Advances in Nuclear Power Plants, May 2007, Nice, France. American Nuclear Society, in2p3-00135149, 7186 (2007).

X. X. Li, Y. W. Ma, C. G. Yu et al., Effects of fuel salt composition on fuel salt temperature coefficient (FSTC) for an under-moderated molten salt reactor (MSR). Nucl. Sci. Tech. 29, 110 (2018). doi: 10.1007/s41365-018-0458-1http://doi.org/10.1007/s41365-018-0458-1

J. R. Keiser, J. H. DeVan, D. L. Manning, The corrosion resistance of type 316 staniess stell to Li2BeF4. Oak Ridge National Lab., ORNL-TM-5782, (1977). doi: 10.2172/7110792http://doi.org/10.2172/7110792

C. W. Lau, C. Demazière, H. Nylén, et al. Improvement of LWR thermal margins by introducing thorium. Prog. Nucl. Energy. 61, 48 (2012). doi: 10.1016/j.pnucene.2012.07.004http://doi.org/10.1016/j.pnucene.2012.07.004

R. L. Murray, K. E. Holbert. Neutron chain reactions - Nuclear Energy (Seventh Edition) - Chapter 16. Nuclear Energy, 16, 259 (2015). doi: 10.1016/B978-0-12-416654-7.00016-2http://doi.org/10.1016/B978-0-12-416654-7.00016-2

James J. Duderstadt, Louis J. Hamilton, Nuclear Reactor Analysis. John Wiley and Sons, first edition, 1976. doi: 10.1109/TNS.1977.4329141http://doi.org/10.1109/TNS.1977.4329141

V. Barkauskas, R. Plukiene, A. Plukis. Actinide-only and full burn-up credit in criticality assessment of RBMK-1500 spent nuclear fuel storage cask using axial burn-up profile. Nucl. Eng. Des. 307, 197 (2016). doi: 10.1016/j.nucengdes.2016.07.012http://doi.org/10.1016/j.nucengdes.2016.07.012

Reactivity Effects Due to Temperature Changes. Nuclear Training Centre, Nuclear Theory, Course 427.00-12, OPG, (1990).

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Influence of 235U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor

DOI：10.1007/s41365-019-0694-z

摘要

Abstract

关键词

Keywords

references

Influence of ²³⁵U enrichment on the moderator temperature coefficient of reactivity in a graphite-moderated molten salt reactor