Analysis of Neutron Fission Reaction Rate in the Nuclear Fuel Cell Using Collision Probability Method with Non Flat Flux Approach

Shafii, Mohammad Ali

doi:10.1088/1742-6596/877/1/012013

Cited by 4 publications

(4 citation statements)

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“…Continuation of fission in the reactor core requires a higher fuel fraction of more extensive fissile materials. This is because, in the fast reactor, the fission reaction rate pertains only to the fuel region 7 . Figure 2 also shows that the k eff of the (U‐Pu)Zr fuel indicates that it is the best option for the SFR core design using the modified CANDLE burn‐up strategy.…”

Section: Resultsmentioning

confidence: 97%

“…Neutronic review is the most crucial aspect in nuclear reactor design and simulates the actual neutron behavior inside the reactor core. The behavior of the neutron, as they react with materials, can be expressed as the neutron flux distribution, whose mathematical representation is given by the neutron transport equation 7,8 . Neutron transport involves the distribution, movement, and interaction of the neutron with the nucleus in the reactor core.…”

Section: Introductionmentioning

confidence: 99%

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Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

et al. 2021

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Summary This study compares three fuel types using neutronic analysis for use in a sodium‐cooled fast reactor (SFR) design with a modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) radial shuffling strategy. SFR is one type of generation IV reactor that is currently under investigation for commercial implementation. In this study, the SFR design utilizes natural uranium as the fuel input. The designed reactor core has a two‐dimensional cylindrical geometry for each fuel mixed oxide (MOX), uranium‐plutonium nitride ([U‐Pu]N), and uranium‐plutonium zirconium ([U‐Pu]Zr). A radial shuffling strategy, using natural uranium as the fuel input, is applied to the SFR to manage the nuclear fuel burn‐up process of the long‐life reactor. This strategy is called the modified CANDLE burn‐up scheme. The reactor core is divided into 10 regions with equal volume to represent the 10 years that the reactor operates without refueling. Initially, the first (innermost) region of the reactor core is filled with natural uranium fuel. The result of the burn‐up from the first region is then shuffled into the second region. The third region is the result of the burn‐up that is shuffled from the second region, and so on. This mechanism only requires natural uranium as the input for each 10‐year fuel cycle. In this study, the fuel movement scheme is examined for three fuels. Global neutronic parameters, such as the multiplication factors and burn‐up analyses, are observed and optimized. Overall, for an output power of 500 MWth and an active core radius of 110 cm and height of 210 cm, the study indicates that (U‐Pu)N is the optimal fuel to be applied in the SFR with a modified CANDLE burn‐up scheme. (U‐Pu)Zr reaches a critical condition at an output power of 500 MWth with an active core radius of 130 cm and a height of 210 cm; whereas criticality for MOX is achieved at an output power of 1500 MWth with an active core radius of 210 cm and a height of 210 cm.

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Section: Resultsmentioning

confidence: 97%

Section: Introductionmentioning

confidence: 99%

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

et al. 2021

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“…The fundamental problem in the nuclear reactor analysis is solving the neutron transport equation. The neutron distribution system is described by the integral transport equation [1]. The integral transport equation is obtained from the integro-differential equation by using some basic assuming.…”

Section: Introductionmentioning

confidence: 99%

“…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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Neutron flux distribution calculation for various spatial mesh of finite slab geometry using one-dimensional diffusion equation

Shafii

Yunanda

Fitriyani

et al. 2019

AIP Conference Proceedings

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Analysis of Neutron Fission Reaction Rate in the Nuclear Fuel Cell Using Collision Probability Method with Non Flat Flux Approach

Cited by 4 publications

References 4 publications

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

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

Neutron flux distribution calculation for various spatial mesh of finite slab geometry using one-dimensional diffusion equation

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Analysis of Neutron Fission Reaction Rate in the Nuclear Fuel Cell Using Collision Probability Method with Non Flat Flux Approach

Cited by 4 publications

References 4 publications

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

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

Neutron flux distribution calculation for various spatial mesh of finite slab geometry using one-dimensional diffusion equation

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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