Thermal-hydraulic Analysis in Wire-wrapped Fuel Assembly for China Lead-based Research Reactor

GE Zeng-fang; ZHOU Tao; BAI Yun-qing; SONG Yong

doi:10.7538/yzk.2015.49.S0.0167

Atomic Energy Science and Technology > 2015 > 49(zengkan1): 167-173. > DOI: 10.7538/yzk.2015.49.S0.0167

GE Zeng-fang, ZHOU Tao, BAI Yun-qing, SONG Yong. Thermal-hydraulic Analysis in Wire-wrapped Fuel Assembly for China Lead-based Research Reactor[J]. Atomic Energy Science and Technology, 2015, 49(zengkan1): 167-173. DOI: 10.7538/yzk.2015.49.S0.0167

Citation:

PDF (5678 KB)

Thermal-hydraulic Analysis in Wire-wrapped Fuel Assembly for China Lead-based Research Reactor

Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei 230031, China

More Information

Graphical Abstract

Abstract

Abstract

In order to reduce the mechanical vibration of fuel rods during the operation and conduce to the coolant mixing between sub-channels, the fuel assembly of lead-based research reactor was designed by using wire-wrapped fixed structure. While the wire-wrapped structure will influence the coolant flow and heat transfer characteristics inside fuel assembly, it’s essential to analyze the influence of wire-wrapped structure on the flow and convective heat transfer in the fuel assembly for optimization design of CLEAR-Ⅰ fuel assembly. In this contribution, fields of velocity, temperature and other flow field characteristics in the wire-wrapped fuel assembly were analyzed using CFD tools. The results show that wire-wraps play an important role on stirring the coolant, and the strength of stirring in inner channel is stronger than in out channel. The friction factor decreases in the entrance region and then attains periodic fluctuations in the fully developed region, and the Nusselt number changes have a same behavior. The cladding highest temperature is under the limit of safety design.
- lead-based research reactor,
- wire-wrap,
- fuel assembly,
- CFD numerical analysis

FullText(HTML)

References (30)

References

[1]	A technology roadmap for generation Ⅳ technology roadmap[R]. US: US DOE Nuclear Energy Research Advisory Committee (NERAC) and the Generation Ⅳ International Forum (GIF), 2002.
[2]	Technology roadmap update for generation Ⅳ nuclear energy systems[R]. Paris: OECD Nuclear Energy Agency for the Generation Ⅳ International Forum, 2014.
[3]	詹文龙，徐瑚珊. 未来先进核裂变能——ADS嬗变系统[J]. 中国科学院院刊，2012，27(3)：375-381.ZHANG Wenlong, XU Hushan. Advanced fission energy program: ADS transmutation system[J]. Bulletin of Chinese Academy of Sciences, 2012, 27(3): 375-381(in Chinese).
[4]	HUANG Q Y, BALUC N, DAI Y, et al. Recent progress of R&D activities on reduced activation ferritic/martensitic steels[J]. Journal of Nuclear Materials, 2013, 442(1-3): 2-8.
[5]	HUANG Q Y, GAO S, ZHU Z Q, et al.Progress in compatibility experiments on lithium-lead with candidate structural materials for fusion in China[J]. Fusion Engineering and Design, 2009, 84: 242-246.
[6]	WU Y C, FDS Team. Design status and development strategy of China liquid lithium-lead blankets and related material technology[J]. Journal of Nuclear Materials, 2007, 367-370: 1410-1415.
[7]	WU Y C, FDS Team. Design analysis of the China dual-functional lithium lead (DFLL) test blanket module in ITER[J]. Fusion Engineering and Design, 2007, 82: 1893-1903.
[8]	WU Y C, HUANG Q Y, ZHU Z Q, et al. R&D of dragon series lithium lead loops for material and blanket technology testing[J]. Fusion Science and Technology, 2012, 62(1): 272-275.
[9]	WU Y C, JIANG J Q, WANG M H, et al. A fusion-driven subcritical system concept based on viable technologies[J]. Nuclear Fusion, 2011, 51(10): 103036.
[10]	WU Y C, SONG G, CAO R F, et al. Development of accurate/advanced radiotherapy treatment planning and quality assurance system (ARTS)[J]. Chinese Physics C, 2008, 32(Suppl. Ⅱ): 177-182.
[11]	WU Y C, FDS Team. CAD-based interface programs for fusion neutron transport simulation[J]. Fusion Engineering and Design, 2009, 84: 1987-1992.
[12]	吴宜灿，柏云清，宋勇，等. 中国铅基研究反应堆概念设计研究[J]. 核科学与工程，2014，34(2)：201-208.WU Yican, BAI Yunqing, SONG Yong, et al. Conceptual design of China Lead-based Research Reactor CLEAR-Ⅰ[J]. Nuclear Science and Engineering, 2014, 34(2): 201-208(in Chinese).
[13]	TENCHINE D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nuclear Engineering and Design, 2010, 240: 1195-1217.
[14]	ROELOFS F, GOPALA V R. Review of fuel assembly and pool thermal hydraulics for fast reactors[J]. Nuclear Engineering and Design, 2013, 265: 1205-1222.
[15]	GAJAPATHY R, VELUSAMY K, SELVARAJ P, et al. CFD investigation of helical wire-wrapped 7-pin fuel bundle and the challenges in modelling full scale 217 bundle[J]. Nuclear Engineering and Design, 2007, 237: 2332-2342.
[16]	刘一哲，喻宏. 中国实验快堆单盒燃料组件内冷却剂流场温度场数值模拟[J]. 原子能科学技术，2007，41（增刊）：230-234.LIU Yizhe, YU Hong. Numerical simulation of flow and temperature field of fuel subassembly for China Experimental Fast Reactor[J]. Atomic Energy Science and Technology, 2007, 41(Suppl.): 230-234(in Chinese).
[17]	刘一哲. 快堆燃料组件热工流体力学计算研究[J]. 原子能科学技术，2008，42(2)：128-134.LIU Yizhe. Thermal hydraulic analysis of fuel subassemblies for sodium-cooled fast reactor[J]. Atomic Energy Science and Technology, 2008, 42(2): 128-134(in Chinese).
[18]	RAZA W, KIM K Y. Comparative analysis of flow and convective heat transfer between 7-pin and 19-pin wire-wrapped fuel -assemblies[J]. Journal of Nuclear Science and Technology, 2008, 45(7): 653-661.
[19]	SREENIVASULU T. Flow and heat transfer characteristics in an annulus wrapped with a helical wire[J]. International Journal of Thermal Sciences, 2009, 48: 1377-1391.
[20]	GAJAPATHY R, VELUSAMY K. A comparative CFD investigation of helical wire-wrapped 7, 19 and 37 fuel pin bundles and its extendibility to 217 pin bundle[J]. Nuclear Engineering and Design, 2009, 239: 2279-2292.
[21]	NATESAN K, SUNDARARAJAN T. Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR[J]. Nuclear Engineering and Design, 2010, 240: 1063-1072.
[22]	PENIGUEL C, RUPP I. Thermal-hydraulics and conjugate heat transfer calculation in a wire-wrapped SFR assembly[C]∥Proceedings of ICAPP 10. San Diego: American Nuclear Society, 2010: 1516-1525.
[23]	BIEDER U, BARTHEL V. CFD calculations of wire-wrapped fuel bundles: Modeling and validation strategies[R]. Bethesda: CFD For Nuclear Reactor Safety Applications (CFD4NRS-3) Workshop, 2010.
[24]	HAMMAN K D, BERRY R A. A CFD simulation process for fast reactor fuel assemblies[J]. Nuclear Engineering and Design, 2010, 240: 2304-2312.
[25]	SREENIVASULU T. Flow and heat transfer characteristics in a seven tube-bundle wrapped with helical wires[J]. International Journal of Advancements in Technology, 2011, 2(3): 350-381.
[26]	李峥. 绕丝组件内流动与传热数值模拟[D]. 哈尔滨:哈尔滨工程大学，2013.
[27]	MENTER F R. Two-equation eddy-viscosity turbulence models for engineering applications[J]. AIAA J, 1994, 32: 269-289.
[28]	CHENG X, TAK N. Investigation on turbulent heat transfer to lead-bismuth eutectic flows in circular tubes for nuclear applications[J]. Nuclear Engineering and Design, 2006, 236: 385-393.
[29]	HAMMAN K D, BERRY R A. A CFD simulation process for fast reactor fuel assemblies[J]. Nuclear Engineering and Design, 2010, 240: 2304-2312.
[30]	FAZIO C, BYUNG CHAN N, LATGE C, et al. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies[M]. Paris: Materials Compatibility, Thermal-hydraulics and Technologies, 2007.

Cited By

Get Citation

PDF

XML

Article views (262) PDF downloads (1229)

Turn off MathJax

Article Contents

Abstract

References

Thermal-hydraulic Analysis in Wire-wrapped Fuel Assembly for China Lead-based Research Reactor

Abstract

References

Catalog

Export File

Citation

Format

Content