The Impact of Neutron Cross Section Group Structures on the Accuracy of Radiological Source Models

Hodgson, A. P. J.; Grimes, Robin W.; Rushton, Michael; Marsden, Olivia J.

doi:10.13182/nse14-156

Cited by 6 publications

(4 citation statements)

References 6 publications

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“…The differences relative to the MCNP results were consistent in comparison with literature with three main exceptions, 89 Zr, 57 Ni, and 56 Mn productions. The primary reason for the difference is the nonoptimal weighting functions for the generation of the multigroup cross sections described in Sections III-B and III-E and highlighted in [36]. The underestimation of the 89 Zr and 57 Ni production cross sections by the SCALE Sampler 252-group simulation results from the 1/E flux weighting of the 13.8-14.6-MeV group utilized.…”

Section: Resultsmentioning

confidence: 99%

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Nuclear Data Covariance Analysis in Radiation-Transport Simulations Utilizing SCALE Sampler and the IRDFF Nuclear Data Library

Quartemont

Bickley

Bevins

2020

IEEE Trans. Nucl. Sci.

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This article describes the nuclear data covariance analysis of an experimental design for a neutron energy-tuning assembly (ETA) created to shape a 14-MeV neutron point source to an objective spectrum. Underlying nuclear data uncertainties play a large role in the radiation transport and reaction rates for the range of responses to be expected from an experiment. The methodology leveraged the Standardized Computer Analysis for Licensing Evaluation (SCALE) Sampler module to determine the uncertainty in the neutron transport. The reaction uncertainty was perturbed with the International Reactor Dosimetry and Fusion File v.1.05 uncertainty, correlation matrix, and reaction cross section through multivariate normal distribution sampling to provide a final response metric. The resultant neutron fluence uncertainty for the ETA ranged from 2.7% to 6.2% in the energy range from 1.28 keV to 14.1 MeV, which contains 99.99% of the neutron fluence. The integrated uncertainties, including statistical and systematic nuclear data uncertainties, for the reaction products analyzed were 2.33% to 4.84% for most reactions, but 55 Mn(n, γ), a less well-characterized reaction occurring in an energy domain with high flux uncertainty, was 19.7%. The mean of the reaction distributions was within 1.1% of the unperturbed nuclear data simulation. The experiment is planned for late 2019, where the predicted results will be compared against the experimental outcomes. The methodology presented can be utilized with alternate nuclear libraries in SCALE to develop uncertainty bounds and neutron flux spectra for many radiation-transport problems.

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Section: Resultsmentioning

confidence: 99%

“…Finally, a study evaluated the activation of materials with a neutron source between various group structures [24]. The reported differences ranged between approximately 5% and 50%, which was directly attributed to the generation of the cross section [36].…”

Section: A Nuclear Data Librariesmentioning

confidence: 99%

Nuclear Data Covariance Analysis in Radiation-Transport Simulations Utilizing SCALE Sampler and the IRDFF Nuclear Data Library

Quartemont

Bickley

Bevins

2020

IEEE Trans. Nucl. Sci.

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“…The energy-dependent flux was calculated using the 172 energy groups in the XMAS structure [9]. The XMAS structure was used to generate the energy-dependent flux spectrum at the D6 NW3 position.…”

Section: Calculation Of Energy-dependent Fluxmentioning

confidence: 99%

Evaluation of neutron radiation damage on gold microstructure using MCNP

Temaugee,

Bedhesi,

Mavunda

et al. 2023

J. Phys.: Conf. Ser.

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The effects of radiation interaction with materials have been studied over the years on metals, semiconductors, and other alloys. The result of these interactions constitutes microstructural effects, like point defects, dislocation loops, and void swellings. The accumulation of these defects results to damage on the macroscopic scale. This study is aimed to predict the magnitude of radiation damage in gold sample due to neutron irradiation. Neutron flux, displacement-per-atom rate, as well as heat deposition, were calculated in MCNP6.2, using the SAFARI-1 reactor model. The total neutron flux and dpa rate in the gold sample were determined to be 2.262 × 1011 n.cm−2.s−1 and 5.209 × 10−7 s−1 respectively. Also, the total heat deposition due to neutrons and photons was 2.515 × 10−6 W.g−1 and 0.513 W.g−1, respectively. Hence, the predicted neutron dpa and flux for the gold sample in this study suggest a heavy damage regime.

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“…Recent studies [10,11] have demonstrated that the fine CCFE-709 and a fortiori the UKAEA-1102 group structure have been optimised to comprehensively cover all applications.…”

Section: Namementioning

confidence: 99%

From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

2017

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Abstract. The nuclear research and development community has a history of using both integral and differential experiments to support accurate lattice-reactor, nuclear reactor criticality and shielding simulations, as well as verification and validation efforts of cross sections and emitted particle spectra. An important aspect to this type of analysis is the proper consideration of the contribution of the neutron spectrum in its entirety, with correct propagation of uncertainties and standard deviations derived from Monte Carlo simulations, to the local and total uncertainty in the simulated reactions rates (RRs), which usually only apply to one application at a time. This paper identifies deficiencies in the traditional treatment, and discusses correct handling of the RR uncertainty quantification and propagation, including details of the cross section components in the RR uncertainty estimates, which are verified for relevant applications. The methodology that rigorously captures the spectral shift and cross section contributions to the uncertainty in the RR are discussed with quantified examples that demonstrate the importance of the proper treatment of the spectrum profile and cross section contributions to the uncertainty in the RR and subsequent response functions.The recently developed inventory code FISPACT-II, when connected to the processed nuclear data libraries TENDL-2015, ENDF/B-VII.1, JENDL-4.0u or JEFF-3.2, forms an enhanced multi-physics platform providing a wide variety of advanced simulation methods for modelling activation, transmutation, burnup protocols and simulating radiation damage sources terms. The system has extended cutting-edge nuclear data forms, uncertainty quantification and propagation methods, which have been the subject of recent integral and differential, fission, fusion and accelerators validation efforts. The simulation system is used to accurately and predictively probe, understand and underpin a modern and sustainable understanding of the nuclear physics that is so important for many areas of science and technology; advanced fission and fuel systems, magnetic and inertial confinement fusion, high energy, accelerator physics, medical application, isotope production, earth exploration, astrophysics and homeland security.

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The Impact of Neutron Cross Section Group Structures on the Accuracy of Radiological Source Models

Cited by 6 publications

References 6 publications

Nuclear Data Covariance Analysis in Radiation-Transport Simulations Utilizing SCALE Sampler and the IRDFF Nuclear Data Library

Nuclear Data Covariance Analysis in Radiation-Transport Simulations Utilizing SCALE Sampler and the IRDFF Nuclear Data Library

Evaluation of neutron radiation damage on gold microstructure using MCNP

From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

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