Three-Dimensional Variational Nodal Transport Methods for Cartesian, Triangular, and Hexagonal Criticality Calculations

Carrico, C.B.; Lewis, E.E.; Palmiotti, G.

doi:10.13182/nse92-1

Cited by 67 publications

(19 citation statements)

References 6 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…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

View full text Add to dashboard Cite

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.

show abstract

Section: Introductionmentioning

confidence: 99%

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

et al. 2012

View full text Add to dashboard Cite

show abstract

“…The geometry consists of three regions as shown in Figure 7, the cross sections for which are given in Table 1. Two configurations of the boundary conditions were considered, and the reference solution was obtained by using the VARIANT code [3][4][5]. For the first configuration, reflected boundary conditions were applied on all sides of the domain, and the reference eigenvalue solution of 0.90152 was obtained with VARIANT (P 17 with sixth-order flux/source and quadratic interface).…”

Section: Ii6 Resultsmentioning

confidence: 99%

Status report on high fidelity reactor simulation.

Palmiotti¹,

Smith²,

Rabiti³

et al. 2006

View full text Add to dashboard Cite

This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and highperformance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermalhydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool Results reported in the AFCI series of technical memoranda frequently are preliminary and subject to revision. Consequently, they should not be quoted or referenced without the authors' permission.

show abstract

“…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

View full text Add to dashboard Cite

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.

show abstract

Three-Dimensional Variational Nodal Transport Methods for Cartesian, Triangular, and Hexagonal Criticality Calculations

Cited by 67 publications

References 6 publications

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

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

Status report on high fidelity reactor simulation.

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

Contact Info

Product

Resources

About