REGAL International Program: Analysis of experimental data for depletion code validation

Eysermans, J.; Verwerft, Marc; Govers, Kevin; Ichou, R.; Ilas, Germina; Meryturek, U.; Messaoudi, Nadia; Romojaro, P.; Slosse, Nicolas

doi:10.1016/j.anucene.2022.109057

Cited by 6 publications

(3 citation statements)

References 4 publications

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“…Very limited code-to-code studies exist that show comparisons between the SCALE 6 releases and other depletion codes. Comparison of code predictions to high-quality experimental data [104] for one PWR UO2 fuel sample shows that nuclide concentration predictions by SCALE 6.2.2/ENDF-B/VII.1 are in good agreement with the measured values for most of the measured nuclides, and these predictions are similar to nuclide concentrations calculated with other depletion codes. For example, the differences between the SCALE/TRITON isotopic inventory predictions and measurement values varied from 0.3% for 235 U to −12.5% for 241 Pu.…”

Section: Depletion Code Validationssupporting

confidence: 59%

Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses

Radulescu¹

2023

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Section: Depletion Code Validationssupporting

confidence: 59%

Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses

Radulescu¹

2023

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“…From a life cycle point of view, reactor operation comes first and criticality safety considerations and the determination of the effective multiplication factor k eff were traditionally of higher priority compared to parameters like neutron and gamma sources for backend activities. Therefore, codes like SCALE or CASMO/SIMULATE have been extensively validated (Rhodes et al, 2009;Saylor et al, 2018;Bahadir, 2020;Eysemans et al, 2022) with regard to factors important for the chain reaction. In particular, this means that cross sections and fission yields of relevant nuclides have been analyzed by dedicated cross section measurements [for example, (Buckner et al, 2017)], by radiochemical analysis [for example, (Govers et al, 2015)] or by criticality experiments [for example, (Sanchez et al, 2021)].…”

Section: Figurementioning

confidence: 99%

Note on the potential to increase the accuracy of source term calculations for spent nuclear fuel

Seidl¹,

Schillebeeckx

Rochman

2023

Front. Energy Res.

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The accuracy of source term predictions is an important factor which determines the efficiency of interim and final storage of spent nuclear fuel. To minimize the required number of storage containers and to minimize the volume and mass of facilities while maintaining safety margins requires accurate codes to predict the decay heat and the gamma and neutron sources with minimum bias for time points ranging from months to thousands of years. While the relevant nuclear data for the purpose of criticality safety received high attention in the last decades and have been extensively verified with dedicated tests, nuclear data relevant for spent nuclear fuel had smaller priority. Mostly results from a radiochemical analysis of samples taken from commercially irradiated fuel have been used for validation purposes. The comparatively sparse data available from tests which exclusively focus on nuclide vector validation under research conditions means that many factors enter the uncertainty estimate of the measurement-theory comparisons and limits the ability to validate codes to a high accuracy. Firstly, the current status of validation efforts is reviewed. Secondly, fields of improvement are explored. Thirdly, the character of uncertainty distributions in measurement-theory comparisons (C/E) of nuclide vectors is analyzed. Currently there are indications that the C/E data is thick tailed which limits improvement of code validation efforts.

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“…The Post-Irradiation Examination (PIE) data are quite useful to validate the evaluated nuclear data, and the PIE data which have been accumulated in the world so far have been utilized in benchmarking the newly-developed evaluated nuclear data files. Recently, a new experimental program, REGAL (the Rod-Extremity and Gadolinia AnaLysis), has been proposed and launched [1]. In this program, the PIE data are acquired for the Gd-bearing fuel rods during (or after) the initial Gd burnout.…”

Section: Introductionmentioning

confidence: 99%

Uncertainty quantification calculations of nuclide number densities of Gd-bearing fuel rods in light water reactors

Chiba

2023

EPJ Web Conf.

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The Post-Irradiation Examination (PIE) data are quite useful to validate the evaluated nuclear data. Recently, a new experimental program, REGAL (the Rod-Extremity and Gadolinia AnaLysis), has been proposed and launched. During this program, the PIE data are acquired for the Gd-bearing fuel rods during (or after) the initial Gd burnout. These PIE data would be different from those for the normal UOX or MOX fuel rods from a viewpoint of nuclide transmutation process. In the present work, we attempt to quantify this difference through calculating sensitivity coefficients of nuclide number densities of the Gd-bearing rods using the depletion perturbation theory and performing uncertainty quantification calculations with the sensitivity coefficients and the nuclear data covariance data. Through the numerical analyses, we found that the impact of the uncertainties of the nuclear data of Gd isotopes on the number densities of actinoids is negligible. The nuclear data-induced uncertainties of Gd isotopes number densities are relatively large during the Gd depletion, and those become small after the Gd burnout. After the Gd burnout, the relative standard deviations of Gd-155 and -157 number densities are approximately 6% and 13%, respectively, and the nuclear data of the parent nuclides of these Gd isotopes, Gd-154 and -156, are also important as well as those of Gd-155 and -157.

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REGAL International Program: Analysis of experimental data for depletion code validation

Cited by 6 publications

References 4 publications

Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses

Updated Recommendations Related to Spent Fuel Transport and Dry Storage Shielding Analyses

Note on the potential to increase the accuracy of source term calculations for spent nuclear fuel

Uncertainty quantification calculations of nuclide number densities of Gd-bearing fuel rods in light water reactors

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