Reactor Fuel Isotopics and Code Validation for Nuclear Applications

Francis, Matthew W.; Weber, C.F.; Pigni, Marco T.; Gauld, Ian C

doi:10.2172/1185693

Cited by 11 publications

(16 citation statements)

References 27 publications

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“…10) does not correspond to the experimental values of [6]. Analyzing other studies performed in this field [7] it was found that possible concentrations of 241 Am in other reactor types show similar results, therefore it is possible that the experimental values are either incorrect or are not presented correctly. Assuming Pu concentrations for different enrichment RBMK-1500 fuel calculated using generated neutron cross-section libraries in constant flux and constant power cases and experimental points from [6] as well as point depletion modelling results from [4].…”

Section: Americium and Neptuniummentioning

confidence: 89%

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generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

et al. 2017

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Depletion of RBMK-1500 spent nuclear fuel (SNF) with and without an erbium burnable absorber was modelled, and onegroup burn-up dependent cross-section libraries for Origen-ARP were created. Depletion calculations for the generation of crosssection libraries were performed using the SCALE 6.1 code package with the TRITON control module, which employs the NEWT deterministic 2D transport code with the 238-group energy library based on the ENDF-B VII library and the ORIGEN-S nuclide composition calculation code. Concentrations of the most important actinides for criticality safety were calculated using the created libraries and were compared with the available experimental data and the newest modelling results. Available experimental data of fission products (Nd and Cs isotopes) were also compared to the modelling results. Composition differences were evaluated for several fuel enrichments and water densities. The comparison shows an acceptable agreement between the values obtained using new one-group cross-section libraries and experimental data except for 238 Pu and 241 Am, as well as the causes of discrepancy are discussed. It has been found that the enrichment and presence of the burnable absorber play an important role in the SNF composition. At the highest evaluated burn-up (29 GWd/tU) isotopic composition differences between 2% enrichment fuel and 2.8% burn-up for actinides important to burn-up credit (BUC) applications varied from 11 to 52%.

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Section: Americium and Neptuniummentioning

confidence: 89%

“…Therefore, computer modelling is often used to evaluate SNF characteristics [2][3][4]. However, validation of these results is also essential: results of calculations of various complexity should correspond to available experimental knowledge [5][6][7].…”

Section: Introductionmentioning

confidence: 99%

generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

et al. 2017

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“…The ORIGEN2 fuel performance code has been experimentally validated for PWR fuel systems and can be considered to provide accurate predictions of SNF compositions. 14,15 All actinide and fission product isotopes where nuclide concentrations are greater than 1.0000 g/tHM at discharge were utilised. The concentration values and their variation with burnup (0-60 GWd/tHM range) presented in the source data are given to five significant figures and were used as recorded.…”

Section: 1: Software and Data Sourcesmentioning

confidence: 99%

“…1, 3,4 Similarly, there are very few complete datasets of spent nuclear fuel compositions, 2,5 though several simplified [6][7][8][9][10][11] and some incomplete sets have been published, 12,13 as part of experimental code validations. 14,15 This means that such data is often beyond the reach of many researchers, serving to hamper research in the development of advanced nuclear technologies.…”

Section: : Introductionmentioning

confidence: 99%

“…A range of nuclides not accessible directly via fission processes due to being "shielded" by stable or long-lived lower-Z elements of the same isotones are produced via neutron capture reactions from lighter FP isotopes during nuclear fuel irradiation. 14 produced from stable 95 Mo. Figure 5 presents an example where the quadratic regression provides an acceptable fit to the source data (R 2 = 0.9991), but a more accurate fit can be gained by the addition of a cubic term (R 2 = 1).…”

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confidence: 99%

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Accessible Statistical Regression, Extrapolation and First-Principles Modelling of Nuclear Data for Spent Nuclear Fuel Composition and Decay Heat Calculations in Short-Term Storage and Recycling

Holdsworth¹,

Sharrad²,

George³

et al. 2020

Preprint

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Computational methods are essential to support and advance nuclear technologies due to the hazards of handling and analysing highly radioactive materials such as spent nuclear fuel (SNF). However, many such methods, including those thatcan predict SNF compositions and decay heat parameters, require expensive, proprietary software, alongside significant programming experience and computational power for utilisation, severely limiting availability of data and hampering research throughput. Although some datasets are available, many are incomplete or only cover certain fuel systems for older reactor types. Research investigating new methods for SNF recycling, for example, requires compositional and decay heat data for fuel systems not covered by extant data, though analogous source data may be available. With this in mind, we have developed a simple, accessible, and flexible method for extrapolation of isotopic, elemental, and decayheat compositions for SNF at discharge and following decay storage before recycling, based on an extant dataset. This semi-empirical method uses physical and mathematical first principles and can be performed using software accessible to all researchers. This provides outputs accurate to within 1% of reference values interpolated within the range of available data for isotopic compositions, with sensible extrapolations at higher burnups beyond those reported, withoverall elemental outputs accurate to within 0.1%of expected totals. In this publication, we present the developmental methodology, some sample data, the present limitations, and options for future development and expansion of functionality.

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Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat

Rochman

Vasiliev

Ferroukhi

et al. 2021

EPJ Nuclear Sci. Technol.

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This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources (nuclear data, operating conditions and manufacturing tolerances) are also provided, and are combined with biases into expanded uncertainties. This study is similar to a previous one on the GU1 sample and fit in the framework of code validation, as well as in the estimation of code predictive power for spent fuel characterization.

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Reactor Fuel Isotopics and Code Validation for Nuclear Applications

Cited by 11 publications

References 27 publications

generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

generation of RBMK-1500 spent nuclear fuel one-group cross-section libraries and their evaluation against experimental data

Accessible Statistical Regression, Extrapolation and First-Principles Modelling of Nuclear Data for Spent Nuclear Fuel Composition and Decay Heat Calculations in Short-Term Storage and Recycling

Analysis for the ARIANE GU3 sample: nuclide inventory and decay heat

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