New aspects in the implementation of the quasi-static method for the solution of neutron diffusion problems in the framework of a nodal method

Caron, Dominic; Dulla, Sandra; Ravetto, Piero

doi:10.1016/j.anucene.2015.02.035

Cited by 13 publications

(3 citation statements)

References 11 publications

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“…Meanwhile, these discrete directions are paralleled based on OpenMP to improve efficiency. Last, a prediction-correction strategy of the quasi-static method (PCQM) [19] is applied for the temporal discretization to accommodate the stiffness in transient.…”

Section: Methods and Modelsmentioning

confidence: 99%

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Zheng

et al. 2017

Int J Energy Res

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Summary Partitioning and transmutation of the minor actinides (MAs) from nuclear power plants are of importance for the nuclear‐energy sustainable development. Fast reactors are applied to transmutation for the hard neutron spectrum and high neutron flux. However, the safety‐related neutronic parameters will become worse when large amounts of MAs are loaded. In this paper, transients of a 600 MWe sodium‐cooled fast reactor for MA transmutation are analyzed by using neutron transport simulation. The control rod withdrawal transients are calculated. Two cases are compared to investigate the influence of loading MAs into the fast reactor core. One is the common core loaded with mixed oxide (MOX) fuel, and the other one is the transmutation core loaded with MOX fuel and MAs. The results indicate that in order to apply similar operation criterion with the common core, the transmutation core with 6% weight fraction of MAs should be operated with more than 30% power reduction. In addition, the results of the transport‐based transient analysis and the point kinetics transient analysis are compared. There are noticeable differences between them, which indicate that the usual way based on the point kinetics calculation is not suitable well for simulating the control rod introduced transients. Copyright © 2017 John Wiley & Sons, Ltd.

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Section: Methods and Modelsmentioning

confidence: 99%

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Zheng

et al. 2017

Int J Energy Res

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“…(1) are modified by dividing the macroscopic cross section for fission neutron production by the effective multiplication eigenvalue, k eff . These equations are solved by means of a nodal discretisation in space [5] and a quasi-static discretisation in time [6]. The macroscopic cross sections characterising the materials present in each homogeneous node are functions of time, either explicitly because of the variation in time of the materials themselves or implicitly because of the variation in time of the temperature of the materials.…”

Section: The Frenetic Codementioning

confidence: 99%

Full-core coupled neutronic/thermal-hydraulic modelling of the EBR-II SHRT-45R transient

et al. 2016

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SUMMARYDuring the last decade the European activities in the field of nuclear fission research include the design of fast reactors cooled by liquid metals. Within this framework, the Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code is being developed at Politecnico di Torino over the last few years. It implements a full-core coupled neutronic/thermalhydraulic model of a liquid-metal-cooled fast reactor as relevant for two of the six options currently under study within the framework of the Generation-IV International Forum, namely the lead-cooled fast reactors and the sodium-cooled fast reactors. The code validation process involves the participation in a coordinated research project of the International Atomic Energy Agency, aiming at testing different computational tools against the shutdown heat removal tests performed many years ago in the sodium-cooled Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory, USA. In this paper, results of the FRENETIC analysis of one of the transients considered in the project, the unprotected EBR-II shutdown heat removal test SHRT-45R, are presented and compared to the measurements, providing the first validation of the coupled neutronic/thermal-hydraulic features of the FRENETIC code.

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“…The de sign and safety as sess ment of new nu clear sys tems sug gested for fu ture im prove ment need ac curate sim u la tion of the be hav ior of neu tron dis tri bu tion dur ing typ i cal op er a tional and ac ci den tal con di tions [1]. Gen er ally, two dif fer ent strat e gies have been usually used for ob tain ing neu tron flux dis tri bu tion in a nuclear re ac tor core in clud ing the de ter min is tic meth ods, e. g. the nodal method [2], and the proba bil is tic methods, e. g. the Monte Carlo method [3,4].…”

Section: Introductionmentioning

confidence: 99%

Three-dimensional time-dependent neutron diffusion simulation using average current nodal expansion method

2020

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In this work, the average current nodal expansion method was developed for the time-dependent neutronic simulation of transients in a nuclear reactor's core. For this purpose, an adopted iterative algorithm was proposed for solving the 3-D time-dependent neutron diffusion equation. In the average current nodal expansion method, the domain of the reactor core can be modeled by coarse meshes for neutronic calculation associated with reasonable precision of results. The discretization of time differential terms in the time-dependent equations was fulfilled, according to the implicit scheme. The proposed strategy was implemented in some kinetic problems including an infinite slab reactor, TWIGL 2-D seed-blanket reactor, and 3-D LMW LWR. At first, the steady-state solution was carried out for each test case, and then, the dynamic neutronic calculation was performed during the time for a specified transient scenario. Obtained results of static and dynamic solutions were verified in comparison with well-known references. Results can indicate the ability of the developed calculator to simulate transients in a nuclear reactor's core.

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New aspects in the implementation of the quasi-static method for the solution of neutron diffusion problems in the framework of a nodal method

Cited by 13 publications

References 11 publications

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Assessment of transient characteristics of fast reactors and influences of minor actinides using neutron transport method

Full-core coupled neutronic/thermal-hydraulic modelling of the EBR-II SHRT-45R transient

Three-dimensional time-dependent neutron diffusion simulation using average current nodal expansion method

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