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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2019.1
- 수록면
- 1,871 - 1,885 (15page)
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The deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three-dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided that the code should be upgraded for the analysis of fast neutron spectrum reactors. This paper presents a coupled code - TULIP/STREAM, developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a two-dimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power dis-tributions were compared with the reference solutions obtained by the continuous energy Monte-Carlo code MCS. With the developed code, a sensitivity study of the number of energy groups, the order of anisotropic PN scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives a keff error smaller than 200 pcm and the root mean square errors of the pin-wise power distributions within 2%.
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참고문헌 (23)
참고문헌 신청
N. Z. Cho / 2002 / Fusion of method of characteristics and nodal method for 3-d whole core transport calculation / Trans. Am. Nucl. Soc 86:322~324
J. Y. Cho / 2013 / Three dimensional nuclear analysis system DeCART/CHORUS/MASTER / ANS Annual Meeting
Y. S. Jung / 2013 / Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers / Ann. Nucl. Energy 62:357~374
B. Collins / 2016 / Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT / J. Comput. Phys 326:612~628
J. Chen / 2018 / A new high-fidelity neutronics code NECP-X / Ann. Nucl. Energy 116:417~428
J. Y. Cho / 2013 / Three dimensional nuclear analysis system DeCART/CHORUS/MASTER / ANS Annual Meeting
Y. S. Jung / 2013 / Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers / Ann. Nucl. Energy 62:357~374
B. Collins / 2016 / Stability and accuracy of 3D neutron transport simulations using the 2D/1D method in MPACT / J. Comput. Phys 326:612~628
J. Chen / 2018 / A new high-fidelity neutronics code NECP-X / Ann. Nucl. Energy 116:417~428
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