Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 149 / Number 2 / Pages 175-188
Akio Yamamoto, Tsutomu Ikeno
Nuclear Technology / Volume 149 / Number 2 / Pages 175-188
Format:electronic copy (download)
In this paper, the effect of a pin-by-pin thermal-hydraulic feedback treatment on the core characteristics at a steady-state condition is investigated using a three-dimensional fine-mesh core calculation code. Currently, advanced nodal codes treat the inside of an assembly as homogeneous, and the temperature distribution inside a node is usually ignored. Namely, the fuel temperature is estimated from the assembly average power density, and the moderator temperature is calculated from the nodewise closed-channel model. However, the validity of a flat temperature distribution inside a node has not yet been investigated, because a three-dimensional pin-by-pin whole-core calculation must be done for comparison. A three-dimensional pin-by-pin nodal-transport code for a pressurized water reactor (PWR) core analysis, SCOPE2, was used in this study since it can directly treat the pin-by-pin feedback effect. A whole-core subchannel analysis code was developed to enhance the thermal-hydraulic capability of SCOPE2. The pin-by-pin feedback models for fuel and moderator temperature were established, and their impact on the core characteristics was investigated in a 3 × 3 multiassembly and the whole PWR core geometries. The calculations showed that modeling of the pin-by-pin temperature distribution revealed a negligible effect on core reactivity and only a slight impact on the radial peaking factor. The difference in the radial peaking factor that is exposed by the pin-by-pin temperature modeling is less than 0.005 in the test calculations.
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