Variational Nodal Methods for Neutron Transport

Dilber, I.; Lewis, E.E.

doi:10.13182/nse85-a27436

Cited by 50 publications

(19 citation statements)

References 4 publications

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“…The 1968 energy groups effective XSs have been collapsed to 33 groups XSs for the flux calculation. The neutron flux distribution has been calculated using the TGV/VARIANT code embedded in the ERANOS code [28,29].…”

Section: Introductionmentioning

confidence: 99%

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

et al. 2012

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Abstract:Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium) during a core disruptive accident (CDA) may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF) accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

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Section: Introductionmentioning

confidence: 99%

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

et al. 2012

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“…The 33 energy groups effective neutron-cross sections (XSs) have been processed with the European Cell Code (ECCO) [7], the JEFF3.1 nuclear data library [8] being employed, and neutron transport calculations have been performed by means of the VARIANT code [9,10]. The models are based on the 1500 MWth French ASTRID CFV (Coeur à Faible effet de Vide, i.e.…”

Section: Assessment Of the Astrid-like Modelsmentioning

confidence: 99%

ASTRID-like Fast Reactor Cores for Burning Plutonium and Minor Actinides

et al. 2015

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A reduction of nuclear waste by transmutation of trans-uranium elements (TRUs), such as Pu and Minor Actinides (MAs) contained in Spent Nuclear Fuel (SNF), is a goal for future reactors. In general, countries with on-going nuclear scenarios would profit from MA mass stabilization, while transmutation of Pu and MAs from SNF could be desired in countries in nuclear phaseout. Both missions can be accomplished by employing fast reactors loaded with fuels containing different amounts of Pu and MAs in a closed (or partially closed) fuel cycle. In this paper, two 1200 MWth sodium-cooled fast reactor cores, based on the French ASTRID design, are proposed for burning TRUs (phase-out option) or only MAs (on-going option). Main attention is focused on the safety and on the transmutation performance. The coolant void effect, in the region including the core and the plenum above and the Doppler constant of both systems are negative also with irradiated fuel. The conversion ratios (CR) of the Pu and MA burners are in the ranges from 0.6 to 0.7 and from 0.5 to 0.6, respectively. These results show a large safety and transmutation potential of ASTRID type reactors.

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“…In this case it is meaningless to increase degree p over 6, because the errors from cross section homogenization are dominant to the results. 7 Results of the Ulchin-1, cycle-1 core analysis (pFEMl)…”

Section: Vol 38 No 4 April 2001mentioning

confidence: 99%

“…Galerkin weighted residual method was used to obtain partial currents in both levels. Dilber and Lewis( 7 ) developed variational nodal method with response matrix for diffusion and transport problems. They introduced the first term of the flux expansion coefficients and test functions be node average flux, and used a functional that guarantees the satisfaction of nodal balance.…”

Section: Introductionmentioning

confidence: 99%

Thep-Version of the Finite Element Method for the Solution of Two-Dimensional Neutron Diffusion Equations

Park

Cho

2001

Journal of Nuclear Science and Technology

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Based on the hierarchical structure, two types of the p-version of the finite element code are developed for solving the two-dimensional neutron diffusion equations. One is the conventional p-type FEM, and the reactor domain is discretized into finite elements, and each element matrices are assembled and solved. In the other, with the domain decomposition approach, the reactor domain is decomposed into subdomains and each subdomain is solved independently. They are coupled by use of incoming and outgoing partial currents along the interfaces, which are obtained analytically from fluxes. Based on the multi-p V-cycle method, multilevel acceleration is implemented into both of the methods. The methods were tested on the IAEA-2D benchmark problem, the initial core of Ulchin Unit 1, and a MOX-fuel loaded core. It turns out that the methods can generate accurate core multiplication factor and assembly average quantities.

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Variational Nodal Methods for Neutron Transport

Cited by 50 publications

References 4 publications

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

ASTRID-like Fast Reactor Cores for Burning Plutonium and Minor Actinides

Thep-Version of the Finite Element Method for the Solution of Two-Dimensional Neutron Diffusion Equations

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