Analysis for the ARIANE GU1 sample: Nuclide inventory and decay heat

Rochman, D.; Vasiliev, A.; Ferroukhi, H.; Hursin, Mathieu

doi:10.1016/j.anucene.2021.108359

Cited by 12 publications

(13 citation statements)

References 27 publications

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“…This can be seen in Table A. 3, where the expanded uncertainties, taking into account the observed biases, are significantly larger than the uncertainties obtained from input variations. It was not the case for the PWR analyzed samples, possibly indicating that the modeling conditions seem more adequate for these samples.…”

Section: Beyond Uncertaintiesmentioning

confidence: 84%

“…This way, uncertainties are obtained at the same time for all quantities calculated by CASMO5: nuclide concentrations as well as decay heat. Apart for the void fraction, all input parameters and their uncertainties were varied in a similar way (same method, same initial variations) for the PWR samples called GU1, GU3 and BM1, presented in references [3][4][5], allowing a consistent comparison of uncertainties.…”

Section: Uncertaintiesmentioning

confidence: 99%

“…The present work is also part of the validation of specific calculation scheme, representing the modelization of a specific irradiation environment (for instance a two-or three-dimensional assembly), the irradiation history (cycle parameters, cooling time), boundary conditions and finally the simulation tool (including neutronic transport, depletion and decay). Recently, a consistent method was applied to the Pressurized Water Rector (PWR) UO 2 samples GU1 and GU3 and to the PWR MOX samples BM1 and BM3 [3][4][5] (similar hypothesis with respect to the simulation codes and methods, nuclear data libraries, uncertainties), allowing to compare the prediction capabilities, as well as calculated uncertainties.…”

Section: Introductionmentioning

confidence: 99%

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Analysis of ENRESA BWR samples: nuclide inventory and decay heat

Rochman

Vasiliev

Ferroukhi

et al. 2022

EPJ Nuclear Sci. Technol.

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In this paper the isotopic compositions from 8 Boiling Water Reactor samples are analyzed following different irradiation assumptions as well as different simulation tools. These samples are part of a proprietary experimental program by a Spanish consortium, and they were obtained from a GE14 assembly irradiated in Sweden. Calculated nuclide concentrations are compared with measured ones providing biases for a selection of isotopes and samples; calculated uncertainties are also provided. Finally, the decay heat from one the sample segment is calculated and compared among the different simulation assumptions. It is shown that depending on the considered nuclear data library and modeling, different contributors affect the calculated quantities, indicating a certain level of prediction power.

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Section: Beyond Uncertaintiesmentioning

confidence: 84%

Section: Uncertaintiesmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Analysis of ENRESA BWR samples: nuclide inventory and decay heat

Rochman

Vasiliev

Ferroukhi

et al. 2022

EPJ Nuclear Sci. Technol.

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“…Uncertainty in the calculated nuclide concentration that is due to uncertainty in nuclear data used in calculations can be important, depending on the considered nuclide, the approach used for nuclear data uncertainty propagation, and the nuclear data and associated covariances. The assessment of this uncertainty is beyond the scope of the current study, and relevant examples can be identified in the published literature [22][23][24].…”

Section: Discussionmentioning

confidence: 99%

Nuclide Inventory Benchmark for BWR Spent Nuclear Fuel: Challenges in Evaluation of Modeling Data Assumptions and Uncertainties

Mertyurek

Ilas

2022

JNE

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This work discusses challenges and approaches to uncertainty analyses associated with the development of a nuclide inventory benchmark for fuel irradiated in a boiling water reactor. The benchmark under consideration is being developed based on experimental data from the SFCOMPO international database. The focus herein is on how to address missing data in fuel design and operating conditions that are important for adequately simulating the time-dependent changes in fuel during irradiation in the reactor. The effects of modeling assumptions and uncertainties in modeling parameters on the calculated nuclide inventory were analyzed and quantified through computational models developed using capabilities in the SCALE code system. Particular attention was given to the impact of the power history and water coolant density on the calculated nuclide inventory, as well as to the effect of geometry modeling considerations not usually addressed in a nuclide inventory benchmark. These considerations include gap closure, channel bow, and channel corner radius, which do not usually apply to regular reactor operation but are relevant for assessing impacts of potential anomalous operating scenarios.

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“…For decay heat, the calorimetric measurements from CLAB (Swedish Central Intermediate Storage Facility), GE-Morris, and HEDL (Hanford Engineering Development Laboratory) were considered, leading to more than 250 cases. To support the recommendations of this subtask, some studies were already published, see references [3][4][5][6][7][8][9][10][11][12][13][14][15]. A global review of the The aim is to produce NDA systems that can be used to provide experimental data to validate the performance of the codes (including nuclear data) and to verify the declared design and operational history of the assemblies.…”

Section: Sfc Subtask 21: Theoretical Study Of Snf Source Termsmentioning

confidence: 99%

Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo

Sjöland¹,

Christensen²,

Evins³

et al. 2023

EPJ Nuclear Sci. Technol.

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SFC is a work package in Eurad that investigates issues related to the properties of the spent nuclear fuel in the back-end of the nuclear fuel cycle. Decay heat, nuclide inventory, and fuel integrity (mechanical and otherwise), and not least the related uncertainties, are among the primary focal points of SFC. These have very significant importance for the safety and operational aspect of the back-end. One consequence is the operation economy of the back-end, where deeper understanding and quantification allow for significant optimization, meaning that significant parts of the costs can be reduced. In this paper, SFC is described, and examples of results are presented at about half-time of the work package, which will finish in 2024. The DisCo project started in 2017 and finished in November 2021 and was funded under the Horizon 2020 Euratom program. It investigated if the properties of modern fuel types, namely doped fuel, and MOX, cause any significant difference in the dissolution behavior of the fuel matrix compared with standard fuels. Spent nuclear fuel experiments were complemented with studies on model materials as well as the development of models describing the solid state, the dissolution process, and reactive transport in the near field. This research has improved the understanding of processes occurring at the interface between spent nuclear fuel and aqueous solution, such as redox reactions. Overall, the results show that from a long-term fuel matrix dissolution point of view, there is no significant difference between MOX fuel, Cr+Al-doped fuel, and standard fuels.

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Analysis for the ARIANE GU1 sample: Nuclide inventory and decay heat

Cited by 12 publications

References 27 publications

Analysis of ENRESA BWR samples: nuclide inventory and decay heat

Analysis of ENRESA BWR samples: nuclide inventory and decay heat

Nuclide Inventory Benchmark for BWR Spent Nuclear Fuel: Challenges in Evaluation of Modeling Data Assumptions and Uncertainties

Spent nuclear fuel management, characterisation, and dissolution behaviour: progress and achievement from SFC and DisCo

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