Research on Simulation Model of Small Integrated Lead-Bismuth Cooled Reactor
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摘要: 铅铋冷却反应堆在安全性、设计简化、防扩散和经济性方面具有巨大的潜力。本文以一体化小型铅铋冷却反应堆为研究对象，建立了基于四方程漂移流模型的螺旋管直流蒸汽发生器二次侧模型、一回路主冷却系统模型、本构模型和比例-积分-微分（PID）控制模型，开展了铅铋冷却反应堆运行控制特性研究。结果表明，稳态计算结果与设计值符合较好，模型能够准确模拟铅铋冷却反应堆的运行特性；快速变负荷运行工况下系统参数超调较小，能够实现反应堆功率跟随蒸汽流量的快速变化；传热管堵塞对反应堆运行有较大的影响，每堵塞1根传热管，蒸汽流量降低约6.7%。Abstract: Lead-bismuth cooled reactor has great potential in safety, design simplification, proliferation resistance and economic performance. In this paper, the small integrated lead-bismuth cooled reactor is taken as the research object. The helical-coil once-through steam generator (S/G) model based on four-equation drift-flux, the primary coolant system model, the constitutive model and the proportional-integral-differential (PID) control model are established, and the operation control characteristics of the lead-bismuth cooled reactor are studied. The results show that the steady-state calculation results are in good agreement with the design values, and the proposed model can accurately simulate the characteristics of the lead-bismuth cooled reactor. Under the condition of rapid load change, the system parameters overshoot is small, and the reactor power can follow the rapid change of steam flow. The blockage of heat transfer tubes has a great influence on the operation of the reactor, and the steam flow decreases by about 6.7% for each heat transfer tube blocked.
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Key words:
- LBE cooled reactor /
- Simulation model /
- PCS /
- Heat transfer tube blockage
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图 2 传热管堵塞事故工况瞬态过程
Figure 2. Transient Process of Heat Transfer Tube Blockage Accident Condition
表 3 满负荷工况下程序计算值与设计参数的比较
Table 3. Comparison between Code Calculation Values and Design Parameters under Full Load Condition
参数 设计值 程序计算值 相对误差/% 堆芯功率/MW 1.500 1.498 0.13 冷却剂流量/(kg·s−1) 207.70 207.71 0.0048 堆芯入口温度/℃ 300.00 300.04 0.013 堆芯出口温度/℃ 350.00 349.97 0.0086 蒸汽流量/(kg·s−1) 7.270 7.273 0.04 蒸汽温度/℃ 300.0 299.6 0.13 二次侧压降/MPa 0.160 0.157 1.9 稳压器液位/mm 65.00 64.82 0.28 
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[1] ZRODNIKOV A V, TOSHINSKY G I, KOMLEV O G, et al. Nuclear power development in market conditions with use of multi-purpose modular fast reactors SVBR-75/100[J]. Nuclear Engineering and Design, 2006, 236(14-16): 1490-1502. doi: 10.1016/j.nucengdes.2006.04.005 [2] GREENSPAN E, HONG S G, LEE K B, et al. Innovations in the ENHS reactor design and fuel cycle[J]. Progress in Nuclear Energy, 2008, 50(2-6): 129-139. doi: 10.1016/j.pnucene.2007.10.022 [3] TAKAHASHI M, UCHIDA S, HATA K, et al. Pb-Bi-cooled direct contact boiling water small reactor[J]. Progress in Nuclear Energy, 2005, 47(1-4): 190-201. doi: 10.1016/j.pnucene.2005.05.020 [4] CHOI S, HWANG I S, CHO J H, et al. URANUS: Korean lead-bismuth cooled small modular fast reactor activities[C]//ASME 2011 Small Modular Reactors Symposium. Washington: ASME, 2011: 107-112. [5] WU Y C. Design and R&D progress of china lead-based reactor for ADS research facility[J]. Engineering, 2016, 2(1): 124-131. doi: 10.1016/J.ENG.2016.01.023 [6] GUO C, ZHAO P C, DENG J, et al. Safety analysis of small modular natural circulation lead-cooled fast reactor SNCLFR-100 under unprotected transient[J]. Frontiers in Energy Research, 2021, 9: 678939. doi: 10.3389/fenrg.2021.678939 [7] SUVDANTSETSEG E. Neutronics and transient analysis of a small fast reactor cooled with natural circulation of lead: ELECTRA: European lead cooled training reactor[D]. Stockholm: KTH Royal Institute of Technology, 2014. [8] YAN M Y, SEKIMOTO H. Safety analysis of small long life CANDLE fast reactor[J]. Annals of Nuclear Energy, 2008, 35(5): 813-828. doi: 10.1016/j.anucene.2007.09.009 [9] WEI S Y, MA W M, WANG C L, et al. Development and validation of transient thermal-hydraulic evaluation code for a lead-based fast reactor[J]. International Journal of Energy Research, 2021, 45(8): 12215-12233. doi: 10.1002/er.6334 [10] YANG Y P, WANG C L, ZHANG D L, et al. Numerical analysis of liquid metal helical coil once-through tube steam generator[J]. Annals of Nuclear Energy, 2022, 167: 108860. doi: 10.1016/j.anucene.2021.108860 [11] HERNANDEZ C R, GRISHCHENKO D, KUDINOV P, et al. Development of a CFD-based model to simulate loss of flow transients in a small lead-cooled reactor[J]. Nuclear Engineering and Design, 2022, 392: 111773. doi: 10.1016/j.nucengdes.2022.111773 [12] FAZIO C. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies[R]. Paris: OECD, 2015. [13] MIKITYUK K. Heat transfer to liquid metal: Review of data and correlations for tube bundles[J]. Nuclear Engineering and Design, 2009, 239(4): 680-687. doi: 10.1016/j.nucengdes.2008.12.014 [14] CHENG X, BATTA A, CHEN H Y, et al. Turbulent heat transfer to heavy liquid metals in circular tubes[C]//ASME 2004 Heat Transfer/Fluids Engineering Summer Conference. Charlotte: ASME, 2004: 115-125. [15] XIA G L, YUAN Y, PENG M J, et al. Numerical studies of a helical coil once-through steam generator[J]. Annals of Nuclear Energy, 2017, 109: 52-60. doi: 10.1016/j.anucene.2017.05.025 [16] 李精精,周涛,刘梦影,等. 铅铋与水自然循环流动传热比较分析[J]. 核科学与工程,2014, 34(2): 249-256. [17] CHENG S K, TODREAS N E. Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles—Bundle friction factors, subchannel friction factors and mixing parameters[J]. Nuclear Engineering and Design, 1986, 92(2): 227-251. doi: 10.1016/0029-5493(86)90249-9 -
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