Assessment of a heat transfer correlations package for water-cooled plasma-facing components in fusion reactors

Yin, Songtao; Cardella, A.; Abdelmessih, A.H.; Jin, Zi; Bromley, Blair P.

doi:10.1016/0029-5493(94)90338-7

Cited by 11 publications

(2 citation statements)

References 5 publications

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“…Reference [22] presents the results of an assessment of a thermal-hydraulics correlation package recently developed for ITER by AECL, for use in the analysis of heat transfer within the divertor cooling channels. Three single-phase heat transfer correlations were assessed, and these are: Dittus-Boelter, Petukhov, and Sieder-Tate.…”

Section: Steady-state Thermohydraulic Analysismentioning

confidence: 99%

Probabilistic Analysis of Divertor Plate Lifetime in Tokamak Reactors

Golinescu

Kazimi

1994

Fusion Technology

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Probabilistic methods have been successfully used to analyze the risk associated with different existing nuclear power plant designs. Such tools can be also used at earlier stages in the design process. Defining a methodology for generation of the probabilistic lifetime reliability for a specific tokamak in-vessel component, the divertor, is the objective of this thesis. The divertor plates establish an interface between the plasma and the material surface of the tokamak device.The design of the divertor cooling system is a most demanding task since it endures the largest power density during operation. Even more severe consequences would appear under some transient conditions.In the present analysis, the divertor conditions for the technology phase of operation of the International Thermonuclear Experimental Reactor (ITER) as specified at the conceptual design stage are used as the reference design. The methodology developed to analyze the ITER divertor reliability consists of the following steps:1. description of normal operating conditions;2. thermal-hydraulic analysis for normal operation;3. identification and classification of transient events; 4. estimation of frequency of occurrence of transient events; 5. thermal analysis for transient event conditions; 6. defining the failure modes / failure criteria models; 7. assessment and propagation of uncertainties; 8. evaluation of the probability of avoiding failure of the divertor plate. The transient events considered in this work are grouped in two categories as follows:-transients that do not affect the divertor temperature distribution prior to shutdown, such as: auxiliary heating system disturbances, magnet system disturbances, main coolant disturbances, balance of plant disturbances, internal plasma disturbances; -transients that affect the divertor temperature distribution prior to shutdown, such as those on the coolant side: Loss of Coolant Accident (LOCA), Loss of Flow Accident (LOFA), and Loss of Heat Sink (LHS), and overpower (OP) transient on the plasma side. In applying this methodology, one failure mode is assumed to be the predominant failure mode: that of surface material loss due to sputtering, melting and evaporation. Taking no credit for redeposition of that material partially balances the fact that other failure modes are not accounted for. 2 The heat conduction computer code HEATING 7.2 is used for the steady-state and transient thermohydraulic analyses as well as for estimating the material loss during transients. The development of a reliability function requires defining a probability distribution function for the material loss during transients. A second order response surface (Response Surface Methodology) is derived for the material loss as a function of the uncertain parameters. The -uncertainties of the frequency of occurrence of transients and material loss parameters are propagated through the reliability function by a Monte Carlo simulation.Using the limited data available leads to the conclusion that there is a high prob...

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Section: Steady-state Thermohydraulic Analysismentioning

confidence: 99%

Probabilistic Analysis of Divertor Plate Lifetime in Tokamak Reactors

Golinescu

Kazimi

1994

Fusion Technology

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“…Heat transfer coefficients during single phase and boiling were calculated by using standard correlations [3][4]. Finite element analysis was performed by COSMOS [5] and TOPAZ2D [6] codes.…”

Section: Thermal Analysismentioning

confidence: 99%

Thermal hydraulic analysis of the TPX plasma facing components

Baxi

Reis

Redler

et al.

15th IEEE/NPSS Symposium. Fusion Engineering

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T This paper describes the thermal hydraulic analysis performed for the plasma facing c o m p o n e n t s (PFC) of the Tokamak Physics Experiment (TPX). The plasma facing material for TPX is graphite and cooling is paovidcd by water. The type of cooling, the mounting method (i.e.? bolting or brazing) and the structurai material depends on the heat flux anticipated in different areas. Since TPX is a longpulse machine, all analysis was done for steady-state conditions. To reduce the radiation enhanoced sublimation, the armor temperature has to be less than a 1000 C toroidal average with a 1200°C local peak. In addition, there are temperature limits on structural materials and belleville wwhers. In order to reduce the impurity sources to ;he plasma, the rninvmilm wall temperature has to be more than 150 C. TPX plasma facing components analyzed in this study were; inboard and outboard limiters, neutral beam armor, ripple plates and divertors. The heat flux on these components varies from 0.15 MW/m2 for the passivc plate armor to 15 MW/m2 for the divertor tiles.From thermal hydraulic considerations, TPX plasma facing components can be dividcd into five types: 1) graphite armor bolted to water-cooled, corrugatcd titanium structured (for heat flux less than 1.0 MW/m2), 2) graphite armor bolted to water-cooled copper plates,3) graphite a r m o r brazed to water-cooled copper plates, 4) graphite monoblocks with DS-copper tubcs, and 5) graphitc Monoblocks with DS-copper tubes with swirl tube inserts (for heat flux >5 MW/m2).0-7803-1 412-3$04.00019943EEE

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Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

Domalapally

Caro

2017

Heat Mass Transfer

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Assessment of a heat transfer correlations package for water-cooled plasma-facing components in fusion reactors

Cited by 11 publications

References 5 publications

Probabilistic Analysis of Divertor Plate Lifetime in Tokamak Reactors

Probabilistic Analysis of Divertor Plate Lifetime in Tokamak Reactors

Thermal hydraulic analysis of the TPX plasma facing components

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

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