Endre søk
RefereraExporteraLink to record
Permanent link

Direct link
Referera
Referensformat
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Annet format
Fler format
Språk
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Annet språk
Fler språk
Utmatningsformat
  • html
  • text
  • asciidoc
  • rtf
The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
2009 (engelsk)Inngår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, nr 8, s. 860-871Artikkel i tidsskrift (Fagfellevurdert) Published
Abstract [en]

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

sted, utgiver, år, opplag, sider
2009. Vol. 51, nr 8, s. 860-871
Emneord [en]
Severe accident, Melt pool formation, Natural convection, Corium, coolability, Effective convectivity, Validation, natural-convection
HSV kategori
Identifikatorer
URN: urn:nbn:se:kth:diva-18847DOI: 10.1016/j.pnucene.2009.06.001ISI: 000270636000011OAI: oai:DiVA.org:kth-18847DiVA, id: diva2:336894
Merknad
QC 20100525Tilgjengelig fra: 2010-08-05 Laget: 2010-08-05 Sist oppdatert: 2017-12-12bibliografisk kontrollert
Inngår i avhandling
1. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head
Åpne denne publikasjonen i ny fane eller vindu >>The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head
2009 (engelsk)Doktoravhandling, med artikler (Annet vitenskapelig)
Abstract [en]

Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents.  In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment.  The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis.

 

The CFD method, on the one hand, is indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling plays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs.

sted, utgiver, år, opplag, sider
KTH: KTH, 2009. s. xx, 76
Serie
TRITA-FYS, ISSN 0280-316X ; 2009:22
Emneord
light water reactor, hypothetical severe accident, accident progression, accident scenario, core melt pool, heat transfer, turbulent natural convection, heat transfer coefficient, phase change, mushy zone, crust, lower plenum, analytical model, effective convectivity model, CFD simulation
HSV kategori
Identifikatorer
urn:nbn:se:kth:diva-10671 (URN)978-91-7415-381-1 (ISBN)
Disputas
2009-09-02, FA32, AlbaNova University Center, Roslagstullsbacken 21, Stockholm, 10:00 (engelsk)
Opponent
Veileder
Merknad

QC 20100812

Tilgjengelig fra: 2009-06-16 Laget: 2009-06-16 Sist oppdatert: 2015-06-18bibliografisk kontrollert

Open Access i DiVA

Fulltekst mangler i DiVA

Andre lenker

Forlagets fulltekst

Søk i DiVA

Av forfatter/redaktør
Tran, Chi-ThanhDinh, Truc-Nam
Av organisasjonen
I samme tidsskrift
Progress in nuclear energy (New series)

Søk utenfor DiVA

GoogleGoogle Scholar

doi
urn-nbn

Altmetric

doi
urn-nbn
Totalt: 102 treff
RefereraExporteraLink to record
Permanent link

Direct link
Referera
Referensformat
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Annet format
Fler format
Språk
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Annet språk
Fler språk
Utmatningsformat
  • html
  • text
  • asciidoc
  • rtf