Abstract
To achieve the high-fidelity and high-performance neutronics simulation of complex assemblies and whole-core problems, the lattice physics code ALPHA (Advanced Lattice Physics code based on Heterogeneous Architecture) is developed. ALPHA employs the planar method of characteristic (MOC) with the axial flux expansion (MOC-EX) scheme to perform 3D problems. The coarse mesh finite difference (CMFD) scheme is used to accelerate the transport sweeping numerically. The heterogeneous system architecture (HSA) has been a powerful tool to accelerate numerous scientific computing applications. Motivated by this, the transport solver of ALPHA is parallelized on the CPU/GPU heterogeneous platform. To exploit the computing capability of the GPU, some optimizations are investigated and implemented on MOC parallel algorithm. In this paper, the implementation, verification, and validation of the neutron transport solver built-in ALPHA are presented.
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References
Wang, B., Liu, Z., Chen, J., Zhao, C., Cao, L., Wu, H.: A modified predictor-corrector quasi-static method in NECP-X for reactor transient analysis based on the 2D/1D transport method. Prog. Nucl. Energy 108, 122–135 (2018). https://doi.org/10.1016/j.pnucene.2018.05.014
Choi, N., Kang, J., Joo, H.G.: Massively parallel method of characteristics neutron transport calculation with anisotropic scattering treatment on GPUs. In: Proceeding of International Conference on High Performance Computing in Asia Pacific Region, Chiyoda, Tokyo, Japan, 28–31 Jan 2018
Choi, N., Kang, J., Lee, H., Joo, H.: Practical acceleration of direct whole-core calculation employing graphics processing units. Prog. Nucl. Energy 133, 103631 (2021)
Song, P., Zhang, Z., Liang, L., et al.: Implementation and performance analysis of the massively parallel method of characteristics based on GPU. Ann. Nucl. Energy 131, 257–272 (2019)
TOP500 official site, 2021. https://www.top500.org/
Zhang, Z., Wang, K., Li, Q.: Accelerating a three-dimensional MOC calculation using GPU with CUDA and two-level GCMFD method. Ann. Nucl. Energy 62, 445–451 (2013)
Zheng, Y.: Study on acceleration techniques of matrix MOC and 3-D neutron transport calculation method. Dissertation of Ph.D., Harbin Engineering University, 2017
Song, P.: Research on the CPU/GPU heterogeneous parallel algorithm for the method of characteristics solution of whole-core neutron transport calculation. Ph.D. thesis, Harbin Engineering University, 2021
Liang, L., Liu, Z., Zheng, Y., Cao, L., Wu, H.: Leakage reconstruction method for 2D/1D fusion transport calculations. Prog. Nucl. Energy 97, 60–73 (2017)
Marin-Lafleche, A., Smith, M.A., Lee, C.: PROTEUS-MOC: a 3D deterministic solver incorporating the 2D method of characteristics. In: M & C 2013: International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, Idaho, USA, 2013
Zhao, C., et al.: Analysis and comparison of the 2D/1D and quasi-3D methods with the direct transport code SHARK. Nucl. Eng. Technol. (2021)
Zheng, Y., Choi, S., Lee, D.: A new approach to three-dimensional neutron transport solution based on the method of characteristics and linear axial approximation. J. Comput. Phys. 350, 25–44 (2017)
Lewis, E.E., Palmiotti, G., Taiwo, T.A., Blomquist, R.N., Smith, M.A., Tsoulfanidis, N.: Benchmark Specifications for Deterministic MOX Fuel Assembly Transport Calculations Without Spatial Homogenization. Organization for Economic Co-operation and Development’s Nuclear Energy Agency (2003)
Yamamoto, A., Tabuchi, M., Sugimura, N., et al.: Derivation of optimum polar angle quadrature set for the method of characteristics based on approximation error for the Bickley function. J. Nucl. Sci. Eng. 44(2), 129–136 (2007)
Romano, P.K., Horelik, N.E., Herman, B.R., Nelson, A.G., Forget, B., Smith, K.: OpenMC: a state-of-the-art Monte Carlo code for research and development. Ann. Nucl. Energy 82, 90–97 (2015)
Acknowledgements
This work is supported by the fund provided by the National Natural Science Foundation of China [12105063], the Science and Technology on Reactor System Design Technology Laboratory [HT-KFKT-02-2019004], the Stability Support Fund for Key Laboratory of Nuclear Data [JCKY2021201C154] and the Natural Science Foundation of Heilongjiang Province of China [LH2020A001].
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Zou, H., Zhang, Q., Song, P., Liang, L., Zhao, Q. (2023). Development of High-Fidelity Neutron Transport Solver of Alpha. In: Liu, C. (eds) Proceedings of the 23rd Pacific Basin Nuclear Conference, Volume 3. Springer Proceedings in Physics, vol 285. Springer, Singapore. https://doi.org/10.1007/978-981-19-8899-8_98
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DOI: https://doi.org/10.1007/978-981-19-8899-8_98
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