Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach

Shafii, Mohammad Ali; Su’ud, Zaki; Waris, Abdul; Kurniasih, Neny

doi:10.4236/wjnst.2012.22008

Cited by 7 publications

(5 citation statements)

References 3 publications

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“…Reactor designs using MOX 20 and (U‐Pu)N 21 fuels have been studied previously, including calculations at the level of nuclear fuel cells 22 . In addition, metallic uranium‐based (U‐Zr) fuels have been previously considered for use in carbon dioxide gas‐cooled fast reactors (GFR) 23 .…”

Section: Design Concept and Methodologymentioning

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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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: Design Concept and Methodologymentioning

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 composition of region number or mesh in each region follows [5]. Specification design of nuclear fuel cell is shown in Table 1.…”

Section: Design and Computational Modelmentioning

confidence: 99%

“…Neutron flux in each region is different each other and flux transformation is governed by linear interpolation scheme [4][5] and quadratic interpolation investigation [6]. As neutron flux is non uniform within the regions of the cell therefore this concept is commonly called as non-flat flux (NFF) approximation [5]. Consequently, the CP matrix is governed not only by the optical path length as exponential parameter but also by the shape of the function that describe the behavior of neutron flux in each region.…”

Section: Introductionmentioning

confidence: 99%

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

Shafii

Fitriyani

Tongkukut

et al. 2017

IOP Conf. Ser.: Mater. Sci. Eng.

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Recent citationsThe Pij matrix and flux calculation of onedimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method Abstract. To solve the integral neutron transport equation using collision probability (CP) method usually requires flat flux (FF) approach. In this research, it has been carried out in the cylindrical nuclear fuel cell with the spatial of mesh with quadratic flux approach. This means that the neutron flux at any region of the nuclear fuel cell is forced to follow the pattern of a quadratic function. The mechanism may be referred to as the process of non-flat flux (NFF) approach. The parameters that calculated in this study are the k-eff and the distribution of neutron flux. The result shows that all parameters are in accordance with the result of SRAC.

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“…Nuclear fuel cell calculation is one of the most complicated steps of neutron transport problems in the reactor core [1]. Broadly speaking, neutron physics problems need to be solved at three levels of precision with respect to the energy variable:  In the epithermal domain, the many resonances of heavy nuclei such as uranium 238 typically require a few tens of thousands of energy groups in order to properly describe each of the resonances.…”

Section: Introductionmentioning

confidence: 99%

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

Hameed¹,

Nagy²,

Abou-Gabal³

2015

Ren. Ener. & Sust. Dev.

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Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time causes difficulties in using any of them as the main calculation code. Monte Carlo codes are used mainly to benchmark the results. Ideally, the flux distribution in the lattice would be determined by solving the transport equation in the exact geometry of the lattice using continuous energy cross sections, the way a Monte Carlo code might. But because of time constraints, the calculation scheme within a lattice physics code is intended to reduce the overall computation time without sacrificing too much accuracy. The nuclear deterministic codes have limited precision due to the approximations made to solve the multi-group transport equation. Self -shielding treatment is responsible for the biggest error in any deterministic code; it is an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor.RESD © 2015 http://apc.aast.edu equivalence and dilution method, and the effect of using Ribbon extended approach on sub-group method.

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Nuclear Fuel Cell Calculation Using Collision Probability Method with Linear Non Flat Flux Approach

Cited by 7 publications

References 3 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

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

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

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