• Title/Summary/Keyword: nuclear fuel rod

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Study on the Quantitative Rod Internal Pressure Design Criterion (정량적인 핵연료봉 내압 설계기준에 관한 연구)

  • Kim, Kyu-Tae;Kim, Oh-Hwan;Han, Hee-Tak
    • Nuclear Engineering and Technology
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    • v.23 no.4
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    • pp.363-373
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    • 1991

  • The current rod internal pressure criterion permits fuel rods to operate with internal pressures in excess of system pressure only if internal overpressure does not cause the diametral gap enlargement. In this study, the generic allowable internal gas pressure not violating this criterion is estimated as a function of rod power. The results show that the generic allowable internal gas pressure decreases linearly with the increase of rod power. Application of the generic allowable internal gas pressure for the rod internal pressure design criterion will result in the simplication of the current design procedure for checking the diametral gap enlargement caused by internal overpressure because according to the current design procedure the cladding creepout rate should be compared with the fuel swelling rate at each axial node at each time step whenever internal pressure exceeds the system pressure.

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DEVELOPMENT OF THE ENIGMA FUEL PERFORMANCE CODE FOR WHOLE CORE ANALYSIS AND DRY STORAGE ASSESSMENTS

  • Rossiter, Glyn
    • Nuclear Engineering and Technology
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    • v.43 no.6
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    • pp.489-498
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    • 2011

  • UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

  • https://doi.org/10.5516/NET.2011.43.6.489 인용 PDF KSCI