Linden code validation against NUPEC / PSBT experimental data for void fraction and temperature benchmarks
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hdl:2117/382747
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Data publicació2022
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Abstract
The subchannel analysis code LINDEN is being developed by China Nuclear Power Technology Research Institute Co. Ltd (CNPRI). The LINDEN code is used in thermal-hydraulics design and safety analysis of pressurized water reactor (PWR) cores. As part of the code development activities, CNPRI commissioned Energy Software Ltd. (ENSO) to conduct an independent assessment of the LINDEN code. The experimental data from Nuclear Power Engineering Corporation (NUPEC) PWR Subchannel and Bundle Tests (PSBT), available through the PSBT benchmark activity, were selected for this validation. The assessment work focused on the void fraction and temperature related benchmarks and was divided into three parts: (1) steady-state void fraction and pressure drop benchmarks, (2) steady-state f luid temperature benchmark, and (3) transient void fraction benchmark. The results presented in this paper correspond to the steady-state parts of the validation work. The assessment of the code comprised the code-to-data comparison as well as the code-to-code comparison. The first one relied on the concepts of accuracy, precision and consistency which can be quantitatively evaluated from statistical indicators and their comparison to the uncertainty of the experimental measurements. The second one consisted in a qualitative assessment against the PSBT benchmark results, with the object of comparing the LINDEN calculated values to other state-of-the-art codes for this type of analysis, and of complementing the code-to-data comparison, which lacked precise information on the experimental uncertainty. The overall conclusion is that the LINDEN calculated values can be considered in good agreement with the PSBT data.
CitacióPérez, M. [et al.]. Linden code validation against NUPEC / PSBT experimental data for void fraction and temperature benchmarks. A: International Topical Meeting on Nuclear Reactor Thermal Hydraulics. "NURETH-19 International Topical Meeting on Nuclear Reactor Thermal Hydraulics". 2022, p. 1-15.
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35684-revised_full_paper.pdf | 734,3Kb | Accés restringit |