• Nuclear Engineering and Technology
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• 제43권6호
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• pp.509-522
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• 2011

FRAPCON-3.4a and FRAPTRAN 1.4 are the most recent versions of the U.S. Nuclear Regulatory Commission (NRC) steady-state and transient fuel performance codes, respectively. These codes have been assessed against separate effects data and integral assessment data and have been determined to provide a best estimate calculation of fuel performance. Recent updates included in FRAPCON-3.4a include updated material properties models, models for new fuel and cladding types, cladding finite element analysis capability, and capability to perform uncertainty analyses and calculate upper tolerance limits for important outputs. Recent updates included in FRAPTRAN 1.4 include: material properties models that are consistent with FRAPCON-3.4a, cladding failure models that are applicable for loss-of coolant-accident and reactivity initiated accident modeling, and updated heat transfer models. This paper briefly describes these code updates and data assessments, highlighting the particularly important improvements and data assessments. This paper also discusses areas of improvements that will be addressed in upcoming code versions.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.980-992
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• 2024

Steam line break accident (SLB) in the nuclear reactor is one of the representative Non-LOCA accidents in which thermal-hydraulics and neutron kinetics are strongly coupled each other. Thus, the multi-scale and multi-physics approach is applied in this study in order to examine a realistic safety margin. An entire reactor coolant system is modelled by system scale node, whereas sub-channel scale resolution is applied for the region of interest such as the reactor core. Fuel performance code is extended to consider full core pin-wise fuel behaviour. The MARU platform is developed for easy integration of the codes to be coupled. An initial stage of the steam line break accident is simulated on the MARU platform. As cold coolant is injected from the cold leg into the reactor pressure vessel, the power increases due to the moderator feedback. Three-dimensional coolant and fuel behaviour are qualitatively visualized for easy comprehension. Moreover, quantitative investigation is added by focusing on the enhancement of safety margin by means of comparing the minimum departure from nucleate boiling ratio (MDNBR). Three factors contributing to the increase of the MDNBR are proposed: Various geometric parameters, realistic power distribution by neutron kinetics code, Radial coolant mixing including sub-channel physics model.