• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.144-153
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• 2003

The object of this paper is the probabilistic failure analysis on the cladding performance of WPF(Whole Pin Furnace) test fuel pins under transient conditions, and analysis of the KALIMER fuel pin using the preceding analysis. The cumulative damage estimation and Weibull probability estimation of WPF test are performed. The probabilistic method was adapted for these analyses to determine the effective thickness thinning due to eutectic penetration depth. In the results, it is difficult to assume that a brittle layer depth made by eutectic reaction is all of the thickness reduction due to cladding thinning. About 93% cladding thinning of the eutectic penetration depth is favorable as an effective thickness of cladding. And the unreliability of the KALIMER driver fuel pin under the same WPF test condition is lower than that of the WPF pin because of the higher plenum-fuel volume ratio and lower cladding inner radius vs. thickness ratio. KALIMER fuel pin developed from conceptual design has a more stable transient performance for a failure mechanism due to fission gas buildup than the WPF pin.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.271-276
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• 1996

Coolant flow distribution to the assemblies and core coolant/component temperatures should be determined in LMR core steady state thermal-hydraulic performance analysis. Sodium flow is distributed to core assemblies with the overall goal of equalizing the peak cladding midwall temperatures for the peak temperature pin of each pin bundle, thus pin cladding damage accrual and pin reliability. The flow orificing analysis for conceptual design will be performed with Excel spreadsheet program ORFCE which was set up and tested, using the calibration factors based on available analyses data. For the verification of this program, flow orificing calculation for the MDP 840MWth core was performed. The calculational results are satisfactory compared to those of CRIEPI calculation.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.284-284
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• 2002

• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.473-478
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• 2004

This work presents the Thermal Hydraulic analysis has been performed for a 19-pin wire-spacer fuel assembly using three-dimensional Reynolds-averaged Navier-Stokes equations. SST model is used as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary condition at inlet and outlet of the calculation domain. The overall results far a preliminary calculation show a good agreement with the experimental observations. It has been found that the major unidirectional flows are the axial velocity in sub-channels and the peripheral sweeping flows and the velocities are found to be following a cyclic path of period equal to the wire-wrap pitch. The temperature is found to be maximum in the central region and also, there exist a radial temperature gradient between the fuel rods. The major advantage of performing this kind of analysis is the prediction of thermal-hydraulic behavior of a fuel assembly with much ease.

• Journal of Energy Engineering
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• v.13 no.3
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• pp.181-190
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• 2004

Computational models for analyzing the in-reactor behavior of metallic fuel pins under transient conditions in liquid-metal reactors are developed and implemented in the TRAMAC (TRAnsient thermo-Mechanical Analysis Code) for a metal fuel rod under transient operation conditions. Not only the basic models for a fuel rod performance but also some sub-models used for transient condition are installed in TRAMAC. Among the models, a fission gas release model, which takes the multi-bubble size distribution into account to characterize the lenticular bubble shape and the saturation condition on the grain boundary and the cladding deformation model have been developed based mainly on the existing models in the MAC-SIS code. Finally, cladding strains are calculated from the amount of thermal creep, irradiation creep, and irradiation swelling. The cladding strain model in TRAMAC predicts well the absolute magnitudes and gen-eral trends of their predictions compared with those of experimental data. TRAMAC results for the FH-1,2,6 pins are more conservative than experimental data and relatively reasonable than those of FPIN2 code. From the calculation results of TRAMAC, it is apparent that the code is capable of predicting fission gas release, and cladding deformation for LMR metal fuel finder transient operation conditions. The results show that in general, the predictions of TRAMAC agree well with the available irradiation data.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.