Benchmarking Of ALEPH2 against MALIBU experimental data

    Research output


    Accurate characterization of nuclear spent fuel is essential for various purposes, including shielding design, fuel licensing, and proper management. While experimental methods offer precise isotopic composition measurements, it is not always economically feasible to measure all nuclides or fuel assemblies. Computational codes, such as the ALEPH2 burnup code developed at SCK CEN, provide an alternative by generating models that closely resemble reality. However, validation of these codes against experimental data is necessary. This research focuses on validating the ALEPH2 code using experiments from the MALIBU Program, specifically analyzing UO2 and MOX fuel samples irradiated until high burnups in the Gösgen Reactor (PWR), Switzerland. The research involves creating a reference model for each sample using provided data and comparing it with experimental measurements. Relative error calculations are performed to quantify the deviations between simulated and experimental values. Additionally, the impact of various parameters, such as temperature, density of moderator, fuel composition, boron evolution with time, and fuel assembly arrangement, is evaluated. The research also includes an assessment of the JENDL-5 and JEFF-4T2 nuclear data libraries, comparing them with other well-known libraries like JEFF-3.3 and ENDF/B-VIII.0. The results from the validation exercise using experiments from the MALIBU program showed significant discrepancies in the burnup levels of the three selected samples, suggesting potential issues with the provided irradiation history. Furthermore, various errors and inconsistencies in the data added an indeterminable bias to the model, making this experiment a less suitable option for code validation, despite ALEPH2 performing adequately. As a result, developing an optimal model for these samples was not feasible. Regarding the benchmarking of nuclear data libraries, significant discrepancies were observed between JENDL-5 and JEFF-3.3 for certain isotopes, including 244Pu and Sm isotopes. The identified sources of discrepancies highlight the need for further research and development in nuclear data evaluation. By including specific files from JEFF-3.3, such as 243Pu neutron transport and 244mAm radioactive decay data for 244Pu, and the 147Pm branching ratio for radiative capture for Sm isotopes, more accurate results were obtained. These additions reduced the relative error between the model and experiment by over 68% for 244Pu, 16% for 148Sm, 10% for 149Sm, and 7% for 150Sm. These findings have been reported to the JENDL-5 developers. Lastly, JEFF-4T2 generally improved the response of the model for numerous isotopes due to alterations in cross-section files rather than fission product yield data. The inconsistencies in the irradiation history, along with various errors and discrepancies found in the MALIBU experiment data, create an unpredictable bias that hinders its suitability for validation exercises. Although ALEPH2 functions properly, it was impossible to propose an optimized model for every sample due to significant deviations between computational models and experimental results. In contrast, the benchmarking of JENDL-5 yields more definitive conclusions. The incorporation of identified files into the JENDL-5 library is recommended based on improvements achieved, which enhances accuracy and reliability. Missing data play a crucial role in concentration calculations and adding them would greatly benefit researchers and scientists relying on JENDL-5. This evaluation offers valuable insights into nuclear data libraries' accuracy and reliability, representing a significant advancement in the field of nuclear data evaluation with room for further improvement.
    Original languageEnglish
    QualificationIndustrial Engineer
    Awarding Institution
    • UPM, Universidad Politécnica de Madrid
    • Cabellos, Oscar, Supervisor, External person
    Date of Award4 Oct 2023
    StatePublished - 4 Oct 2023

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