• NUCLEAR TECHNIQUES
  • Vol. 46, Issue 3, 030601 (2023)
Mingrui YANG, Qizheng SUN, Chixu LUO, Donghao HE, Xiaojing LIU, and Tengfei ZHANG*
Author Affiliations
  • School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
  • show less
    DOI: 10.11889/j.0253-3219.2023.hjs.46.030601 Cite this Article
    Mingrui YANG, Qizheng SUN, Chixu LUO, Donghao HE, Xiaojing LIU, Tengfei ZHANG. Development and verification of a neutronics-thermal hydraulics coupling code with unstructured meshes neutron transport model[J]. NUCLEAR TECHNIQUES, 2023, 46(3): 030601 Copy Citation Text show less
    References

    [1] Zhang H, Zhuang J. Research, development and industrial application of heat pipe technology in China[J]. Applied Thermal Engineering, 23, 1067-1083(2003).

    [2] Snoj L, Trkov A, Jaćimović R et al. Analysis of neutron flux distribution for the validation of computational methods for the optimization of research reactor utilization[J]. Applied Radiation and Isotopes, 69, 136-141(2011).

    [3] Zhang T F, Xiao W, Yin H et al. VITAS: a multi-purpose simulation code for the solution of neutron transport problems based on variational nodal methods[J]. Annals of Nuclear Energy, 178, 109335(2022).

    [4] Yang D M, Liu X J, Zhang T F et al. A comparison of three algorithms applied in thermal-hydraulics and neutronics codes coupling for lbe-cooled fast reactor[J]. Annals of Nuclear Energy, 149, 107789(2020).

    [5] Wang J C, Wang Q, Ding M. Review on neutronic/thermal-hydraulic coupling simulation methods for nuclear reactor analysis[J]. Annals of Nuclear Energy, 137, 107165(2020).

    [6] Zhang T F, Li Z P. Variational nodal methods for neutron transport: 40 years in review[J]. Nuclear Engineering and Technology, 54, 3181-3204(2022).

    [7] Chao Y A, Attard A. Resolution of the stiffness problem of reactor kinetics[J]. Nuclear Science and Engineering, 90, 40-46(1985).

    [8] Shen Q, Wang Y, Jabaay D et al. Transient analysis of C5G7-TD benchmark with MPACT[J]. Annals of Nuclear Energy, 125, 107-120(2019).

    [9] Wang B, Liu Z, Chen J et al. A modified predictor-corrector quasi-static method in NECP-X for reactor transient analysis based on the 2D/1D transport method[J]. Progress in Nuclear Energy, 108, 122-135(2018).

    [10] Xiao W, Sun Q Z, Liu X J et al. Application of stiffness confinement method within variational nodal method for solving time-dependent neutron transport equation[J]. Computer Physics Communications, 279, 108450(2022).

    [11] Křepel J, Rohde U, Grundmann U et al. DYN3D-MSR spatial dynamics code for molten salt reactors[J]. Annals of Nuclear Energy, 34, 449-462(2007).

    [12] Ghiaasiaan S M, Wassel A T, J L jr Farr et al. Heat conduction in nuclear fuel rods[J]. Nuclear Engineering and Design, 85, 89-96(1985).

    [13] NIU Yuhang, HE Yanan, WU Yingwei et al. Analysis of primary loop system of high-order fully-implicit nuclear reactor based on MOOSE platform[J]. Nuclear Power Engineering, 42, 50-57(2021).

    [14] Ban Y, Endo T, Yamamoto A. A unified approach for numerical calculation of space-dependent kinetic equation[J]. Journal of Nuclear Science and Technology, 49, 496-515(2012).

    [15] He M T, Wu H C, Cao L Z et al. Time-dependent, three dimensional nodal transport code development based on unstructured mesh[C](2014).

    [17] Seubert A, Sureda A, Bader J et al. The 3-D time-dependent transport code TORT-TD and its coupling with the 3D thermal-hydraulic code ATTICA3D for HTGR applications[J]. Nuclear Engineering & Design, 251, 173-180(2012).

    [18] Finnemann H, Galati A. NEACRP 3-D LWR core transient benchmark, final specification[R](1991).

    [19] Barber D A, Miller R M, Joo H G et al. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes[R]. Los Alamos National Lab(1999).

    [20] WANG Lianjie, ZHAO Wenbo, CHEN Bingde et al. Development of coupled 3-D neutronics/thermal-hydraulics code for SCWR core transient analysis[J]. Nuclear Power Engineering, 35, 186-189(2014).

    [21] HE Mingtao. Transport-based transient methods of liquid-metal cooled fast reactors and transient characteristics of minor actinides transmutation[D](2016).

    [22] Finnemann H, Bauer H, Galati A et al. Results of LWR core transient benchmarks[R](1993).

    Mingrui YANG, Qizheng SUN, Chixu LUO, Donghao HE, Xiaojing LIU, Tengfei ZHANG. Development and verification of a neutronics-thermal hydraulics coupling code with unstructured meshes neutron transport model[J]. NUCLEAR TECHNIQUES, 2023, 46(3): 030601
    Download Citation