The Investigation of Burnup Characteristics Using the Serpent Monte Carlo Code for a Sodium Cooled Fast Reactor

Korkmaz, Mehmet Emin; Ağar, O.

doi:10.5516/net.07.2013.050

Cited by 6 publications

(2 citation statements)

References 15 publications

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“…The two methods are Transmutation Trajectory Analysis (TTA) and Chebyshev Rational Approximation Method (CRAM). In the present study, the CRAM was used based on its matrix exponential solution (Leppänen, 2009a;Jaakko, 2013;Korkmaz and Agar, 2014;Herrero et al, 2016;Kępisty et al, 2017). Furthermore, the cross-section data libraries: radioactive decay and fission yield and also thermal scattering libraries are all used by the Serpent.…”

Section: Introductionmentioning

confidence: 99%

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

Abubakar¹,

Ismaila²,

Yamusa³

2022

FJS

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The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

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Section: Introductionmentioning

confidence: 99%

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

Abubakar¹,

Ismaila²,

Yamusa³

2022

FJS

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“…The SFR is the most promising reactor type because it can achieve the Generation IV goals in a reasonable amount of time, based on experience accumulated over the years 2 . Generation IV reactors have the potential to ensure safety, economic competitiveness, reduction in environmental burden, efficient utilization of resources, resistance to nuclear proliferation, and enhanced physical protection 3,4 . Recently, design and construction work have started on active advanced SFR research projects, such as the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) in France and Prototype Generation‐IV Sodium‐cooled Fast Reactor (PGSFR) in South Korea 5 .…”

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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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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Directional diffusion coefficients and leakage-corrected discontinuity factors: Implementation in Serpent and tests

Dorval

2016

Annals of Nuclear Energy

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The Investigation of Burnup Characteristics Using the Serpent Monte Carlo Code for a Sodium Cooled Fast Reactor

Cited by 6 publications

References 15 publications

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

Serpent Code Point Kinetic Parameters Analysis: A Nexus for Nigeria Research Reactor – 1 Core Conversion Using Enriched Uranium Dioxide Fuel

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

Directional diffusion coefficients and leakage-corrected discontinuity factors: Implementation in Serpent and tests

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