SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

Michel-Sendis, F.; Gauld, Ian C; Martínez, J.S.; Alejano, C.; Bossant, M.; Boulanger, Damien; Cabellos, Ó.; Chrapciak, V.; Conde, J.M.; Fast, Ivan; Gren, M.; Govers, Kevin; Gysemans, Mireille; Hannstein, Volker; Havluj, Frantisek; Hennebach, M.; Hordósy, G.; Ilas, Germina; Kilger, R.; Mills, Robert W.; Mountford, David; Ortego, Pedro; Radulescu, Georgeta; Rahimi, M.; Ranta-Aho, A.; Rantamäki, Karin; Ruprecht, B.; Soppera, N.; Stuke, Maik; Suyama, Kenya; Tittelbach, S.; Töre, Candan; Winckel, Stefaan Van; Vasiliev, A.; Watanabe, Tomoaki; Yamamoto, Toru; Yamamoto, Tadahiro

doi:10.1016/j.anucene.2017.07.022

Cited by 80 publications

(30 citation statements)

References 10 publications

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“…With a different experimental data set, the FYs of U-235 could be more precisely targeted. The SFCOMPO database [55,56] could be used to find more post-irradiation examination data. The methods presented in this work could also be used to provide high quality adjustments to the FY data.…”

Section: Discussionmentioning

confidence: 99%

Data assimilation of post-irradiation examination data for fission yields from GEF

Siefman

Hursin

Sjöstrand

et al. 2020

EPJ Nuclear Sci. Technol.

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Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using post-irradiation examination experiments. The adjusted fission yields lowered the bias and reduced the uncertainty of the simulations. Our approach adjusts the model parameters of the code GEF. We compare the BFMC and MOCABA approaches to data assimilation, focusing especially on the effects of the non-normality of GEF’s fission yields. In the application that we present, the best data assimilation framework decreased the average bias of the simulations from 26% to 14%. The average relative standard deviation decreased from 21% to 14%. The GEF fission yields after data assimilation agreed better with those in JEFF3.3. For Pu-239 thermal fission, the average relative difference from JEFF3.3 was 16% before data assimilation and after it was 12%. For the standard deviations of the fission yields, GEF’s were 100% larger than JEFF3.3’s before data assimilation and after were only 4% larger. The inconsistency of the integral data had an important effect on MOCABA, as shown with the Marginal Likelihood Optimization method. When the method was not applied, MOCABA’s adjusted fission yields worsened the bias of the simulations by 30%. BFMC showed that it inherently accounted for this inconsistency. Applying Marginal Likelihood Optimization with BFMC gave a 2% lower bias compared to not applying it, but the results were more poorly converged.

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Section: Discussionmentioning

confidence: 99%

Data assimilation of post-irradiation examination data for fission yields from GEF

Siefman

Hursin

Sjöstrand

et al. 2020

EPJ Nuclear Sci. Technol.

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“…e top portion of Figure 1 presents the radial layout of the assembly. In this figure, the location of a special rod is marked, labeled "F6," where samples have been taken for isotopic analysis [21]. e F6 rod has an active height of 371 cm with 4.5 wt.% enrichment.…”

Section: Fukushima Daini-2 2f2d1 Assemblymentioning

confidence: 99%

“…e adjusted sample power history is shown in Figure 4(a). e original power history and other neutronic parameters of this benchmark can be found in the SFCOMPO database [21].…”

Section: Multiphysics Modeling Scale/t5mentioning

confidence: 99%

Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations

Price

Radaideh

Mui

et al. 2020

Science and Technology of Nuclear Installations

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Multiphysics coupling of neutronics/thermal-hydraulics models is essential for accurate modeling of nuclear reactor systems with physics feedback. In this work, SCALE/TRACE coupling is used for neutronic analysis and spent fuel validation of BWR assemblies, which have strong coolant feedback. 3D axial power profiles with coolant feedback are captured in these advanced simulations. The methodology is applied to two BWR assemblies (2F2DN23/SF98 and 2F2D1/F6), discharged from the Fukushima Daini-2 unit. Coupling is performed externally, where the SCALE/T5-DEPL module transfers axial power data in all axial nodes to TRACE, which in turn calculates the coolant density and temperature for each of these nodes. Within a burnup step, the data exchange process is repeated until convergence of all coupling parameters (axial power, coolant density, and coolant temperature) is observed. Analysis of axial power, criticality, and coolant properties at the assembly level is used to verify the coupling process. The 2F2D1/F6 benchmark seems to have insignificant void feedback compared to 2F2DN23/SF98 case, which experiences large power changes during operation. Spent fuel isotopic data are used to validate the coupling methodology, which demonstrated good results for uranium isotopes and satisfactory results for other actinides. This work has a major challenge of lack of documented data to build the coupled models (boundary conditions, control rod history, spatial location in the core, etc.), which encourages more advanced methods to approximate such missing data to achieve better modeling and simulation results.

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“…The expertise, practices, methodologies, and application to support and sustain integral benchmark data [6] initiated via the ICS-BEP and IRPhEP can be applied towards other areas for modernized computational validation. Ongoing activities at the OECD NEA include planned development of a benchmark handbooks supporting SINBAD (Shielding Integral Benchmark Archive and Database) experiment data, [7] the Spent Fuel Composition (SFCOMPO) database, [8]…”

Section: International Benchmark Legacymentioning

confidence: 99%

Current overview of ICSBEP and IRPhEP benchmark evaluation practices

Bess

Ivanova

2020

EPJ Web Conf.

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Two projects sanctioned by the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) have over two decades of experience developing established and comprehensive data sets in handbooks supporting criticality safety and reactor physics. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) serve as examples of quality and excellence in preserving our experimental data heritage and establishing integral benchmark standards upon which current and future modeling, validation, and safety efforts can be supported. Evaluation practices have evolved with each year of these projects to include additional benchmark experiment data, establish more comprehensive techniques for evaluation of uncertainties and biases, and encourage established high-quality peer-review efforts. This paper will summarize the current format of the handbooks, best-practices for a comprehensive benchmark evaluation, recent activities and protocol within these projects, and a look into future identified needs and activities.

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SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

Cited by 80 publications

References 10 publications

Data assimilation of post-irradiation examination data for fission yields from GEF

Data assimilation of post-irradiation examination data for fission yields from GEF

Multiphysics Modeling and Validation of Spent Fuel Isotopics Using Coupled Neutronics/Thermal-Hydraulics Simulations

Current overview of ICSBEP and IRPhEP benchmark evaluation practices

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