Natural circulation characteristics analysis of a small modular natural circulation lead–bismuth eutectic cooled fast reactor

2015 ◽  
Vol 83 ◽  
pp. 220-228 ◽  
Author(s):  
Pengcheng Zhao ◽  
Shuzhou Li ◽  
Zhao Chen ◽  
Jie Zheng ◽  
Hongli Chen
2016 ◽  
Vol 90 ◽  
pp. 54-61 ◽  
Author(s):  
Minghan Yang ◽  
Yong Song ◽  
Jianye Wang ◽  
Peng Xu ◽  
Guangyu Zhang

2000 ◽  
Vol 37 (1-4) ◽  
pp. 211-216 ◽  
Author(s):  
Jong-Eun Chang ◽  
Kune Y. Suh ◽  
Il Soon Hwang

2019 ◽  
Vol 34 (4) ◽  
pp. 313-324
Author(s):  
M. Khizer ◽  
Zhang Yong ◽  
Yang Guowei ◽  
Wu Qingsheng ◽  
Wu Yican

In this study, the structural integrity of liquid metal fast reactor fuel assembly has been established for different parameters considering the optimum fuel design. Analytical calculation of added mass effect due to lead bismuth eutectic and verification through previously presented theories, has been established. The integrity of the hexagonal wrapper of fuel assembly has been guaranteed over the entire operating temperature range. Effect of temperature on the density of lead bismuth eutectic, the subsequent change in added mass of lead bismuth eutectic, the effect on natural frequencies and effect on stresses on wrapper, has been studied in detail. A simple empirical relationship is presented for estimation of added mass effect for lead bismuth eutectic type fast reactors for any desired temperature. An approach for assessment of fast reactor fuel assembly performance has been outlined and calculated results are presented. Nuclear seismic rules require that systems and components which are important to safety, shall be capable of bearing earthquake effects and their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on wrapper and seismic aspect has also been considered using ANSYS. Modal analysis has been compared in vacuum and lead bismuth eutectic using the calculated added mass.


Author(s):  
Hae-Yong Jeong ◽  
Kwi-Seok Ha ◽  
Won-Pyo Chang ◽  
Yong-Bum Lee ◽  
Dohee Hahn ◽  
...  

The Korea Atomic Energy Research Institute (KAERI) is developing a Generation IV sodium-cooled fast reactor design equipped with a passive decay heat removal circuit (PDRC), which is a unique safety system in the design. The performance of the PDRC system is quite important for the safety in a simple system transient and also in an accident condition. In those situations, the heat generated in the core is transported to the ambient atmosphere by natural circulation of the PDRC loop. It is essential to investigate the performance of its heat removal capability through experiments for various operational conditions. Before the main experiments, KAERI is performing numerical studies for an evaluation of the performance of the PDRC system. First, the formation of a stable natural circulation is numerically simulated in a sodium test loop. Further, the performance of its heat removal at a steady state condition and at a transient condition is evaluated with the real design configuration in the KALIMER-600. The MARS-LMR code, which is developed for the system analysis of a liquid metal-cooled fast reactor, is applied to the analysis. In the present study, it is validated that the performance of natural circulation loop is enough to achieve the required passive heat removal for the PDRC. The most optimized modeling methodology is also searched for using various modeling approaches.


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