Neutron Distribution in the Nuclear Fuel Cell using Collision Probability Method with Quadratic Flux Approach

Shafii, Mohammad Ali; Fitriyani, Dian; Tongkukut, Seni H. J.; Abdullah, Ade Gafar

doi:10.1088/1757-899x/180/1/012040

Cited by 2 publications

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References 5 publications

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“…In previous research [1,6,7], the CP method was used to solve the transport of neutrons in a 1D cylindrical nuclear fuel cell. The integral transport is solved using CP method with non flat flux approach which is usually solved by the flat flux approach.…”

Section: Introductionmentioning

confidence: 99%

The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

et al. 2018

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Calculation of Pij matrix of one-dimensional neutron transport in the slab geometry of the nuclear fuel cell using Collision Probability (CP) method has been done. Pij matrix is one of important parameters within the distribution of neutron flux in the nuclear fuel cell. The CP method is the most efficient methods to solve the neutron transport equation in the reactor core. The study is focused on neutron interaction with nuclear fuel cell of U-235 and U-238 for homogeneous condition. The parameters to calculate the Pij matrix are the cross section of nuclear fuel, width of the region and number of regions. A lattice of slabs have been constructed using void boundary conditions for model of finite system to calculate the collision probabilities. If the Pij matrix has been calculated then neutron flux can be determined. The results show that total value of Pij matrix using CP method for U-235 and U-238 is less than one, respectively. This is in accordance with the definition of void boundary conditions for finite slab geometry. Along with Pij matrix, neutron flux is also appropriate with the reference.

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Section: Introductionmentioning

confidence: 99%

The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

et al. 2018

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Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

Shafii

Putri

Handayani

et al. 2021

J. Phys.: Conf. Ser.

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Study of neutronic performance analysis of Sodium-Cooled Fast Reactor (SFR) design based on the variations of output power using radial fuel suffling strategy has been performed. SFR is a type of the generation IV reactor that is widely researched for application on a commercial scale. The reactor design uses natural uranium as a fuel and Sodium as a coolant. The research has been carried out by using the SRAC code and JENDL-3.2 as a library with two dimensional R- Z suffling strategy of cylinder core for variations of power output 300, 350, 400, 450 and 500 MWTh. The neutronic parameters such as a multiplication factor (k-eff and k-inf) and burn-up analysis are observed. The SFR core is separated into 10 regions having the same volume fraction in the radial direction. At beginning of burn-up process, the reactor core is only filled with natural uranium fuel called fresh fuel and prepared for the one cycle of 10 years of burn-up time. The burn-up result in the first region is shifted into the second region, the burn-up result in the second region is shifted into the third region, and so on until burn-up result into the tenth region. The burn-up result in the tenth region is removed from the reactor core, then the first region can be filled with fresh natural uranium fuel and so on up to 10 times fuel cycles of reactor operation. The neutron calculation results indicate that the multiplication factors (k-eff and k-inf) in a critical condition are occurred on the 300 MWTh of output power. Overall, the output power of 300 MWTh has requirements and a greater chance of being operated for SFR designed.

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Neutron Distribution in the Nuclear Fuel Cell using Collision Probability Method with Quadratic Flux Approach

Cited by 2 publications

References 5 publications

The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

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Neutron Distribution in the Nuclear Fuel Cell using Collision Probability Method with Quadratic Flux Approach

Cited by 2 publications

References 5 publications

The Pij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

The Pij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

Study of Neutronic Performance of Sodium-Cooled Fast Reactor Design for Various Output Power Using Radial Fuel Shuffling Strategy

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The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method