Conceptual Design Study of Ultra-high Flux Fast Neutron Research Reactor Fuel
-
摘要: 提高中子注量率是高通量研究堆的发展趋势，能够大幅加速反应堆材料研发进程。但若提高中子注量率至1016 cm−2·s−1将导致功率密度峰值相较于现有研究堆高数倍，对反应堆和核燃料设计带来许多挑战。为此，本文从中子学、传热、燃料材料堆内行为等方面半定量分析了提高中子注量率对核燃料性能的影响，并提出应对超高通量和功率密度挑战的设计措施，为发展超高通量快中子研究堆燃料设计提供指导。Abstract: Increasing neutron fluence rate is the development trend of high flux research reactor, which can greatly accelerate the R&D process of reactor materials. However, if the neutron fluence rate is increased to 1016 cm−2·s−1, the peak power density will be several times higher than that of the existing research reactor, which will bring many challenges to the reactor and nuclear fuel design. For this reason, this paper semi quantitatively analyzes the impact of increasing neutron fluence rate on the performance of nuclear fuel from neutronics, heat transfer, fuel material behavior in the reactor and other aspects, and proposes design measures to meet the challenges of ultra-high flux and power density, providing guidance for the development of ultra-high flux fast neutron research reactor fuel design.
-
Key words:
- Ultra-high flux /
- Research reactor /
- Nuclear fuel /
- Heat transfer /
- Design
-
表 1 研究堆用冷却剂关键参数对比
Table 1. Comparison of Key Parameters of Coolant for Research Reactor
冷却剂参数 钠（723 K，
0.1 MPa）铅铋合金
（723 K，
0.1 MPa）铅（723 K，
0.1 MPa）水（333 K，
2.0 MPa）熔点/K 371 398 600 273 沸点/K 1156 1943 2010 618 $ \rho $/(g·cm−3) 0.84 10.15 10.52 0.98 $ {c}_{p} $/(kJ·kg−1·K−1) 1.300 0.146 0.147 4.180 $ {V}_{\mathrm{m}\mathrm{a}\mathrm{x}} $/(m·s−1) 10.0 4.0 4.0 12.0 ${V}_{\mathrm{m}\mathrm{a}\mathrm{x} } \rho {c}_{p}$/(MW·K−1·m−2) 11.0 5.9 6.2 49.3 $ {t}_{\mathrm{m}\mathrm{a}\mathrm{x}} $/K 823 723 823 373 ${t}_{_{\mathrm{i}\mathrm{n},\mathrm{m}\mathrm{i}\mathrm{n} }}$/K 411 438 640 298 ${t}_{_{\mathrm{m}\mathrm{a}\mathrm{r}\mathrm{g}\mathrm{i}\mathrm{n} }}$/K 336 144 48 58 
下载: 导出CSV
-
[1] CACUCI D G. Handbook of nuclear engineering[M]. New York: Springer, 2010: 1247. [2] 孙博华. 量纲分析与Lie群[M]. 北京: 高等教育出版社, 2016: I. [3] ZHANG J S. A review of steel corrosion by liquid lead and lead-bismuth[J]. Corrosion Science, 2009, 51(6): 1207-1227. doi: 10.1016/j.corsci.2009.03.013 [4] YOSHIDA E, FURUKAWA T. Corrosion issues in sodium-cooled fast reactor (SFR) systems[M]. FÉRON D. Nuclear Corrosion Science and Engineering. Amsterdam: Elsevier, 2012: 773-806. [5] DAI Y N, ZHENG X T, DING P S. Review on sodium corrosion evolution of nuclear-grade 316 stainless steel for sodium-cooled fast reactor applications[J]. Nuclear Engineering and Technology, 2021, 53(11): 3474-3490. doi: 10.1016/j.net.2021.05.021 [6] WANG H, XIAO J, WANG H, et al. Corrosion behavior and surface treatment of cladding materials used in high-temperature lead-bismuth eutectic alloy: A review[J]. Coatings, 2021, 11(3): 364. doi: 10.3390/coatings11030364 [7] OGAWA T, IWAI T, KURATA M. Demixing of U-Zr alloys under a thermal gradient[J]. Journal of the Less Common Metals, 1991, 175(1): 59-69. doi: 10.1016/0022-5088(91)90349-9 [8] LU Y, JIANG Z, LI L Y, et al. Calculation of steady-state dynamical phase diagram in U-Mo binary system under irradiation[J]. Journal of Nuclear Materials, 2021, 544: 152698. doi: 10.1016/j.jnucmat.2020.152698 [9] LU Y, JIANG Z, LI L Y, et al. Calculation of dynamical phase diagram in U-Zr binary system under irradiation[J]. Journal of Nuclear Engineering and Radiation Science, 2021, 7(1): 011605. doi: 10.1115/1.4047718 -
表(1)
计量
- 文章访问数: 345
- HTML全文浏览量: 207
- PDF下载量: 80
- 被引次数: 0
