Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor

Guang-Liang Yang; Hai-Long Liao; Tao Ding; Hong-Li Chen

doi:10.1007/s41365-022-01045-7

Your Location：

Home >

Browse articles >

Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor

NUCLEAR ENERGY SCIENCE AND ENGINEERING | Updated：2022-07-27

- Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor
- Nuclear Science and Techniques Vol. 33, Issue 5, Article number: 66(2022)
- Affiliations：
  
  1.School of Nuclear Science and Technology, University of Science and Technology of China, Hefei, 230026, China
- Author bio：
  
  hlchen1@ustc.edu.cn
- Funds：
- DOI：10.1007/s41365-022-01045-7
  CLC：
Scan for full text
Guang-Liang Yang, Hai-Long Liao, Tao Ding, et al. Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor. [J]. Nuclear Science and Techniques 33(5):66(2022)
DOI：

Guang-Liang Yang, Hai-Long Liao, Tao Ding, et al. Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor. [J]. Nuclear Science and Techniques 33(5):66(2022) DOI： 10.1007/s41365-022-01045-7.

摘要

Abstract

The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors. In this study, considering various irradiation behaviors, a fuel rod performance analysis code, named KMC-Fueltra, was developed to evaluate the thermal-mechanical performance of oxide fuel rods under both normal and transient conditions in the LMFR. The accuracy and reliability of the KMC-Fueltra were validated by analytical solutions as well as the results obtained from codes and experiments. The results indicated that KMC-Fueltra can predict the performance of oxide fuel rods under both normal and transient conditions in the LMFR.

关键词

Keywords

Fuel rod analysis codeThermal-mechanical performanceIrradiation behaviorsPellet-cladding mechanical interactionLiquid metal fast reactor

references

X. Luo, C. Wang, Z.R. Zou et al., Development and application of a multi-physics and multi-scale coupling program for lead-cooled fast reactor. Nucl. Sci. Tech. 33, 18 (2022). doi: 10.1007/s41365-022-01008-yhttp://doi.org/10.1007/s41365-022-01008-y

V.P. Tran, K.C. Nguyen, D. Hartanto et al., Development of a PARCS/Serpent model for neutronics analysis of the Dalat nuclear research reactor. Nucl. Sci. Tech. 32, 15 (2021). doi: 10.1007/s41365-021-00855-5http://doi.org/10.1007/s41365-021-00855-5

W.S. Duan, Z.R. Zou, X. Luo et al., Startup scheme optimization and flow instability of natural circulation lead-cooled fast reactor SNCLFR-100. Nucl. Sci. Tech. 32, 133 (2021). doi: 10.1007/s41365-021-00970-3http://doi.org/10.1007/s41365-021-00970-3

H.Y. Liao, K.L. Lu, H. Yao et al., Development of a transient module based on FROBA-ROD code and assessment for fuel rod performance under reactivity-initiated accident. Ann. Nucl. Energy. 160, 108377 (2021). doi: 10.1016/j.anucene.2021.108377http://doi.org/10.1016/j.anucene.2021.108377

Y.N. He, P. Chen, Y.W. Wu et al., Preliminary evaluation of U3Si2-FeCrAl fuel performance in light water reactors through a multi-physics coupled way. Nucl. Eng. Des. 328, 27-35 (2018). doi: 10.1016/j.nucengdes.2017.12.019http://doi.org/10.1016/j.nucengdes.2017.12.019

R. Godesar, M. Guyette, N. Hoppe, COMETHE II-A computer code for predicting the mechanical and thermal behavior of a fuel pin. Nucl. Appl. Technol. 9(2): 205-217 (1970). doi: 10.13182/NT70-A28809http://doi.org/10.13182/NT70-A28809

V.Z. Jankus, R.W. Weeks, LIFE-II - A computer analysis of fast-reactor fuel-element behavior as a function of reactor operating history. Nucl. Eng. Des. 18(1): 83-96 (1972). doi: 10.1016/0029-5493(72)90038-6http://doi.org/10.1016/0029-5493(72)90038-6

J. Wordsworth, IAMBUS-1 - A digital computer code for the design, in-pile performance prediction and post-irradiation analysis of arbitrary fuel rods. Part I: Theory and modelling[J]. Nucl. Eng. Des. 31(3): 309-336 (1974). doi: 10.1016/0029-5493(75)90167-3http://doi.org/10.1016/0029-5493(75)90167-3

L. Luzzi, A. Cammi, V.D. Marcello et al., Application of the TRANSURANUS code for the fuel pin design process of the ALFRED reactor. Nucl. Eng. Des. 277, 173-187 (2014). doi: 10.1016/j.nucengdes.2014.06.032http://doi.org/10.1016/j.nucengdes.2014.06.032

Y. Tsuboi, H. Endo, T. Ishizu et al., Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code ‘FEMAXI-FBR’. J. Nucl. Sci. Technol. 49, 408-424 (2012). doi: 10.1080/00223131.2012.669242http://doi.org/10.1080/00223131.2012.669242

A. Karahan, and J. Buongiorno, Modeling of thermo-mechanical and irradiation behavior of mixed oxide fuel for sodium fast reactors. J. Nucl. Mater. 396.2-3, 272-282 (2010). doi: 10.1016/j.jnucmat.2009.11.021http://doi.org/10.1016/j.jnucmat.2009.11.021

K.J. Miles and D. J. Hill. DEFORM-4: fuel pin characterization and transient response in the SAS4A accident analysis code system. No. CONF-860501-10. Argonne National Lab., IL (USA), 1986. URL: https://www.osti.gov/biblio/5737020https://www.osti.gov/biblio/5737020

K. Mikityuk, A. Shestopalov, FRED fuel behaviour code: Main models and analysis of Halden IFA-503.2 tests. Nucl. Eng. Des. 241.7, 2455-2461 (2011). doi: 10.1016/j.nucengdes.2011.04.033http://doi.org/10.1016/j.nucengdes.2011.04.033

C. Huang, Master Thesis, Fast reactor fuel rod life analysis program. China Institute of Atomic Energy, 2000. URL:https://xueshu.baidu.com/usercenter/paper/show?paperid=d1191b859359523417f3956ec12f137e&site=xueshu_sehttps://xueshu.baidu.com/usercenter/paper/show?paperid=d1191b859359523417f3956ec12f137e&site=xueshu_se

A.M. Ross, R.L. Stoute, Heat transfer coefficient between UO2 and Zircaloy-2. No. AECL--1552. Atomic Energy of Canada Limited, 1962. URL: https://inis.iaea.org/search/search.aspx?orig_q=RN:43103492https://inis.iaea.org/search/search.aspx?orig_q=RN:43103492

D.D. Lanning, C.R. Hann, Review of methods applicable to the calculation of gap conductance in Zircaloy-clad UO2 fuel rods. No. BNWL-1894. Battelle Pacific Northwest Labs., Richland, Wash.(USA), 1975. URL: https://www.osti.gov/servlets/purl/4209005https://www.osti.gov/servlets/purl/4209005

K. Mikityuk, Heat transfer to liquid metal: review of data and correlations for tube bundles. Nucl. Eng. Des. 239.4, 680-687 (2009). doi: 10.1016/j.nucengdes.2008.12.014http://doi.org/10.1016/j.nucengdes.2008.12.014

A.H. Booth. A method of calculating fission gas diffusion from UO2 fuel and its application to the X-2-f loop test. No. AECL--496. Atomic Energy of Canada Limited, 1957. URL: https://inis.iaea.org/search/search.aspx?orig_q=RN:43123872https://inis.iaea.org/search/search.aspx?orig_q=RN:43123872

J.A. Turnbull, C. A. Friskney, The relation between microstructure and the release of unstable fission products during high temperature irradiation of uranium dioxide. J. Nucl. Mater. 71.2, 238-248 (1978). doi: 10.1016/0022-3115(78)90421-Xhttp://doi.org/10.1016/0022-3115(78)90421-X

M. V. Speight, A calculation on the migration of fission gas in material exhibiting precipitation and re-solution of gas atoms under irradiation. Nucl. Sci. En. 37.2, 180-185 (1969). doi: 10.13182/NSE69-A20676http://doi.org/10.13182/NSE69-A20676

M.H. Wood, J. R. Matthews, A simple operational gas release and swelling model: I. Intragranular gas. J. Nucl. Mater. 91.1, 35-40 (1980). doi: 10.1016/0022-3115(80)90029-Xhttp://doi.org/10.1016/0022-3115(80)90029-X

J.R. Matthews, M.H. Wood, A simple operational gas release and swelling model: II. Grain boundary gas. J. Nucl. Mater. 91.2-3, 241-256 (1980). doi: 10.1016/0022-3115(80)90224-Xhttp://doi.org/10.1016/0022-3115(80)90224-X

F.P. Qi. Development and Application of a Fuel Element Behavior Analysis code for Liquid Metal Fast Reactor. University of Science and Technology of China, 2018. URL: http://cdmd.cnki.com.cn/Article/CDMD-10358-1018104971.htmhttp://cdmd.cnki.com.cn/Article/CDMD-10358-1018104971.htm

V.Z. Jankus, and R. W. Weeks, LIFE-II—A computer analysis of fast-reactor fuel-element behavior as a function of reactor operating history. Nucl. Eng. Des. 18.1, 83-96 (1972). doi: 10.1016/0029-5493(72)90038-6http://doi.org/10.1016/0029-5493(72)90038-6

L.J. Siefken, E.W. Coryell, E.A. Harvego et al. "SCDAP/RELAP5/MOD 3.3 code manual, MATPRO-A library of materials properties for light water reactor accident analysis." Idaho National Engineering and Environmental Laboratory 2001 (2001). URL: https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6150/v4/index.htmlhttps://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6150/v4/index.html

J.D. Hales, R.L. Williamson, S.R. Novascone et al. BISON theory manual the equations behind nuclear fuel analysis. No. INL/EXT-13-29930. Idaho National Lab.(INL), Idaho Falls, ID (United States), 2016. doi: 10.2172/1374503http://doi.org/10.2172/1374503

Sobolev, Vitaly. "Database of thermophysical properties of liquid metal coolants for GEN-IV." (2011). URL:https://publications.sckcen.be/portal/files/356024/Coolants_GEN-IV(BLG-1069_rev1).pdfhttps://publications.sckcen.be/portal/files/356024/Coolants_GEN-IV(BLG-1069_rev1).pdf

A. D’Angelo, B. Arien, V. Sobolev et al. "Benchmark on beam interruptions in an accelerator-driven system. Final report on phase 1 calculations." (2003). URL: https://inis.iaea.org/search/search.aspx?orig_q=RN:34079713https://inis.iaea.org/search/search.aspx?orig_q=RN:34079713

Yang G, Guo Z, Wang P, et al. "Research and validation on the numerical algorithm of mechanical module in a transient fuel rod performance code for fast reactor." Ann. Nucl. Energy. 171 (2022): 108991. doi: 10.1016/j.anucene.2022.108991http://doi.org/10.1016/j.anucene.2022.108991

L.K. Cao, G.L. Yang, H.L. Chen. "Transient sub-channel code development for lead-cooled fast reactor using the second-order upwind scheme." Prog. Nucl. Energy 110 (2019): 199-212. doi: 10.1016/j.pnucene.2018.09.022http://doi.org/10.1016/j.pnucene.2018.09.022

T. Aoyama. The operational experience of the experimental fast reactor'JOYO'. No. JAERI-M--92-028. 1992. URL: https://inis.iaea.org/search/search.aspx?orig_q=RN:23087288https://inis.iaea.org/search/search.aspx?orig_q=RN:23087288

M. Konstantin, S. Pelloni, P. Coddington et al., "FAST: An advanced code system for fast reactor transient analysis." Ann. Nucl. Energy. 32.15 (2005): 1613-1631. doi: 10.1016/j.anucene.2005.06.002http://doi.org/10.1016/j.anucene.2005.06.002

Views

Downloads

CSCD

Alert me when the article has been cited

Submit

Tools

Publicity Resources

No data

Related Author

No data

Related Institution

No data

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