A specialized code for operation transient analysis and its application in fluoride salt-cooled high temperature reactors

Jian Ruan; Bo Xu; Ming-Hai Li; Yang Yang; Yang Zou; Hong-Jie Xu

doi:10.1007/s41365-017-0268-x

Your Location：

Home >

Browse articles >

A specialized code for operation transient analysis and its application in fluoride salt-cooled high temperature reactors

NUCLEAR ENERGY SCIENCE AND ENGINEERING | Updated：2021-02-23

- A specialized code for operation transient analysis and its application in fluoride salt-cooled high temperature reactors
- Nuclear Science and Techniques Vol. 28, Issue 8, Article number: 119(2017)
- Affiliations：
  
  1.Shanghai Institute of Applied Physics, CAS, Shanghai, 201800, China
  2.University of Chinese Academy of Sciences, Beijing 100049, China
- Author bio：
  
  zouyang@sinap.ac.cn;
  Corresponding author E-mail address: xuhongjie@sinap.ac.cn;
- Funds：
- DOI：10.1007/s41365-017-0268-x
  CLC：
Scan for full text
Jian Ruan, Bo Xu, Ming-Hai Li, et al. A specialized code for operation transient analysis and its application in fluoride salt-cooled high temperature reactors. [J]. Nuclear Science and Techniques 28(8):119(2017)
DOI：

Jian Ruan, Bo Xu, Ming-Hai Li, et al. A specialized code for operation transient analysis and its application in fluoride salt-cooled high temperature reactors. [J]. Nuclear Science and Techniques 28(8):119(2017) DOI： 10.1007/s41365-017-0268-x.

摘要

Abstract

Fluoride salt-cooled High-temperature Reactors (FHRs) includemany attractive features, such as high temperature, large heat capacity, low pressureand strong inherent safety. Transient characteristics of FHR areparticularly important for evaluating its operation performance.Thus, a specialized code OCFHR (Operation and Control analysis code of FHR)isused to studyan experimental FHR’s operation behaviors. The geometric modeling of OCFHR is based on one-dimensional lumped parameter method,and some simplifications are taken into consideration during simulation due to the existence of complex structures such as pebble bed, intermediate heat exchanger (IHX),air radiator (AR) and multiply channels. Apoint neutron kinetics modelis developed and neutronphysics calculation is neededto provide some key inputs including axial power density distribution, reactivity coefficients and parameters about delayed neutron precursors. For analyzing the operational performance, five disturbed transients are simulated, involvingreactivity step insertion, variations of coolant mass flow rate of primaryloop andintermediate loop, adjustment of air inlet temperature,and mass flow rate of air-cooling system. Simulation results indicate that inherent self-stabilityof FHRrestrains severe consequencesunderabove transients, and some dynamic featuresareobserved, such as large negative temperature feedbacks, remarkable thermal inertia, and high response delay.

关键词

Keywords

FHRSimulationPebble bedTransient analysis;

references

C. Forsberg, The advanced high-temperature reactor: high-temperature fuel, liquid salt coolant, and liquid-metal reactor plant. Prog. Nucl. Energy 47(1-4), 32-43 (2005). DOI: 10.1016/j.pnucene.2005.05.002http://doi.org/10.1016/j.pnucene.2005.05.002

TMSR center, SINAP. Pre-conceptual Design of a 2MW Pebble Bed Fluoride Salt Coolant High Temperature Test Reactor. (2012).

D.F. Williams, Assessment of CandidateMolten Salt Coolants for theAdvanced High-Temperature Reactor (AHTR). ORNL-TM-2006-12 U.S. Department of energy (2006). DOI: 10.2172/885975.http://doi.org/10.2172/885975.

C.H. Andrades, A.T. Cisneros, J.K. Choi, et al., Technical Description of the “Mark 1” Pebble-Bed Fluoride-salt-Cooled High-Temperature Reactor (PB-FHR) Power Plant. UCBTH-14-002. Department of Nuclear Engineering, University of California, Berkeley (2014). DOI: 10.13140/RG.2.1.4915.7524http://doi.org/10.13140/RG.2.1.4915.7524.

M. Chen. Transient analysis of an FHR coupled to a helium Brayton power cycle. Prog. Nucl. Energy 83, 283-293 (2015). DOI: 10.1016/j.pnucene.2015.02.015http://doi.org/10.1016/j.pnucene.2015.02.015.

M. Li, J. Zhang, Y. Zou et al., Disturbed Transient Analysis with Stable Operation Mode of TMSR-SF1. NURETH-16, Hyatt Regency Chicago, America, August 30 - September 4, on CD-ROM (2015) http://www.ans.org/store/item-700399/http://www.ans.org/store/item-700399/

Q. Lv, H.C. Lin, I.H. Kim et al., DRACS thermal performance evaluation for FHR. Annuals of Nuclear Energy 77,115-128(2015).DOI: 10.1016/j.anucene.2014.10.032http://doi.org/10.1016/j.anucene.2014.10.032.

G. Ablay. A modeling and control approach to advanced nuclear power plants withgas turbines. Energy Conversion and Management 76,899-909(2013). DOI: 10.1016/j.enconman.2013.08.048http://doi.org/10.1016/j.enconman.2013.08.048

C. Forsberg, L.W. Hu, P. Peterson et al. Fluoride-salt-cooled high-Temperature Reactor(FHR) for Power and Process Heat. MIT-ANP-TR-157 (2014).

X. Huang. Research of Runge-Kutta method. Heilongjiang Sci. Tec. 23,23-27(2012). DOI: 10.3969/j.issn.1673-1328.2012.23.063http://doi.org/10.3969/j.issn.1673-1328.2012.23.063 (In Chinese)

M.R. Galvin. System Model of a Natural Circulation Integral Test Facility. Oregon State University (2009).

W. vanAntwerpen et al., A review of correlations to model the packing structure and effective thermalconductivity in packed beds of mono-sized spherical particles.Nuclear Engineering and Design 240, 1803-1818 (2010). DOI: 10.1016/j.nucengdes.2010.03.009http://doi.org/10.1016/j.nucengdes.2010.03.009

N. Wakao et al., Effect of Fluid Dispersion coefficients on Particle-to-Fluid heat transfer coefficients in Packed Beds. Chemical Engineering Sci. 34, 325-336 (1978). DOI: 10.1016/0009-2509(79)85064-2http://doi.org/10.1016/0009-2509(79)85064-2.

S. Ergun. Fluid Flow Through Packed Columns. AIChE J, 18(2):361-371(1972).

S.S. Lomakin, Yu. A. Nechaev. Transient processes and the measurement of reactivity of a reactor containing Beryllium. Translated from AtomnayaEnergiya. Vol 18. No.1, January, 33-40 (1965).DOI: 10.1007/BF01116353http://doi.org/10.1007/BF01116353.

S. Yang et al. (In Chinese) Heat Transfer. BeiJing Higher Education Press: 243-263,(2006).

M. Lin et al., Extension of RELAP5 core power calculation model. Nuclear Power Engineering 28, 16-19(2007). DOI: 10.3969/j.issn.0258-0926.2007.06.005http://doi.org/10.3969/j.issn.0258-0926.2007.06.005

Views

Downloads

CSCD

Alert me when the article has been cited

Submit

Tools

Publicity Resources

On the viability of wearing evaluation by Thin Layer Activation in the presence of non-occupationally exposed individuals

Sensitivity of the area method with mono isotopic fission chambers to reactivity changes in subcritical nuclear reactors

A method for sharing dynamic geometry information in studies on liquid-based detectors

Design and testing of an internal hot-cathode-type PIG ion source for superconducting cyclotron

Assessment of the power deposition on the MEGAPIE spallation target using the GEANT4 toolkit

Related Author

No data

Related Institution

Comissão Nacional de Energia Nuclear (SEASE/DRS/CNEN)

Instituto de Engenharia Nuclear (IEN-CNEN)

Graduate Program of Nuclear Engineering, COPPE, FederalUniversity of Rio de Janeiro

AGH University of Science and Technology, Faculty of Energy and Fuels, al. Mickiewicza 30

Sino-French Institute of Nuclear Engineering and Technology, Sun Yat-sen University

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