TY  - EJOU
AU  - Hao, Dahua 
AU  - Liu, Qiqing 
AU  - Yu, Yin 
AU  - Hu, Yile 

TI  - Study	of	Multi-Group	Neutron	Diffusion	in	Nuclear	Fuel	Pellet	based	on	 Peridynamics
T2  - The International Conference on Computational \& Experimental Engineering and Sciences

PY  - 2023
VL  - 26
IS  - 2
SN  - 1933-2815

AB  - In	 this	 study,	 a	 method	 for	 solving	 multigroup	 neutron	 diffusion	 equations	 for	 nuclear	 fuel	 pellets	 is	
proposed	 based	 on	 the	 bond-based	 PeriDynamic	 (PD)	 theory.	 Firstly,	 adopting	 the	 idea	 of	 non-local	
diffusion,	 the	PD	neutron	diffusion	coefficient	is	defined	and	calibrated	 through	 the	equality	of	potential	
with	 the	 traditional	 neutron	 diffusion	 coefficient.	 Comparing	 the	 calculation	 results	 of	 the	 neutron	 flux	
distribution	 of	 the	 single-group	 neutron	 diffusion	 by	 the	 PD	 method	 and	 the	 traditional	 finite	 element	
method,	the	feasibility	of	the	method	is	verified.	Secondly,	apply	the	leakage	term	in single-group	to	multigroup	and	consider	the	scattering	term	between	different	energy	groups.	Therefore,	the	calculation	of	the	
multi-group	neutron	diffusion	equation	is	 further	 studied.	Through	 the	 research	 on	 the	 four-group	 twodimensional	square	plate	model,	the	reference	value	of	the	effective	multiplication	coefficient	(0.872297)	
and	the	neutron	flux	distribution	results	are	in	good	agreement	with	those	in	the	reference	literature.	Also,	
calculation	model	with	different	materials	is	considered	with	the	parameters	across	the	interface	using	1/2	
of	those	on	two	sides.	Finally,	heat	transfer	caused	by	neutron	diffusion	is	considered,	where	the	heat	source	
is	determined	by	neutron	flux	distribution.	Therefore,	coupling	between	the	neutron	and	thermal	diffusion	
equation	can	be	achieved.
KW  - Peridynamic;	neutron	diffusion	equation;	nuclear	fuel	pellets;	neutron	flux

DO  - 10.32604/icces.2023.09301