A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor

Andersson, Peter; Holcombe, Scott

doi:10.1016/j.anucene.2017.06.025

Cited by 13 publications

(5 citation statements)

References 6 publications

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“…Second, experimental measurements of the power distribution obtained from a specific fuel assembly near the flux trap are used to validate the results obtained from the model. In this study, the power distribution is inferred from the gamma emission intensity of the fission products along the FA [36][37][38][39]. The underlying concept of the experimental measurements is based on the proportionality between the fission power density experienced by the fuel and the concentration of fission products in it.…”

Section: Methodsmentioning

confidence: 99%

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

Abuzlf,

Castagna,

Makmal

et al. 2023

International Journal of Energy Research

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In this study, a high-fidelity Monte Carlo model was utilized to investigate the localized behavior of the power peaking factor in a reactor core. The model considered the effect of water gaps, also known as flux traps, on the local power peaking. The model, developed using the Serpent code, employs a fine spatial mesh to accurately describe the fission power and burnup distribution. The results obtained from the model were validated through comparison with experimental measurements of the power distribution obtained from the IRR1 facility. The study found that the local accumulation of thermal flux caused by enhanced neutron moderation in the flux trap leads to a highly localized increase in power density, affecting the power peaking factor. The quantification of this effect was a key finding of the study. The results obtained from the study were highly dependent on the model’s fidelity, with significant differences being observed between the power peaking factor calculated on a fine mesh as opposed to that calculated at the plate or assembly scale. Furthermore, the experimental validation of the model enabled the prediction of the power peaking factor in other regions within the fuel assembly, which are not accessible to measurement instrumentation, with a high degree of confidence. The study’s findings may be useful for optimizing reactor design and operation and assessing the safety margins in reactors with similar characteristics.

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Section: Methodsmentioning

confidence: 99%

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

Abuzlf,

Castagna,

Makmal

et al. 2023

International Journal of Energy Research

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“…It should be noted that from the simulated spectra one just obtains the probability of full-energy deposition per source particle, therefore, to get the total counts at each position we have multiplied the projection data by a factor such that at the central scan position approximately 1000 counts should be obtained after the multiplication. It was assumed that a minimum of 1000 counts (based on a previous tomographic reconstruction experience of nuclear fuel object [17]) at this position would be required to have a satisfactory reconstruction, also assuming the counts as Poisson distributed, this would result in uncertainty of ± 3%. This was done for all reconstruction cases.…”

Section: Test Object Reconstructionsmentioning

confidence: 99%

Calculation of Spatial Response of a Collimated Segmented HPGe Detector for Gamma Emission Tomography by MCNP Simulations

Rathore

Senis

Jansson

et al. 2022

IEEE Trans. Nucl. Sci.

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We have proposed a planar electronically segmented HPGe detector concept in combination with a multi-slit collimator for gamma emission tomography. In this work, the spatial resolution achievable by using the collimated segmented HPGe detector was evaluated, prior to the manufacture and operation of the detector. The spatial response of a collimated segmented HPGe detector concept was evaluated using simulations performed with Monte Carlo N-Particle transport code MCNP6. The full detector and multi-slit collimator system were modeled and for the quantification of the spatial response, the Modulation Transfer Function (MTF) was chosen as a performance metric. The MTF curve was obtained through the calculation of the Line Spread Function (LSF) by analyzing simulated projection data. In addition, tomographic reconstructions of the simulated simplified test objects were made to demonstrate the performance of the segmented HPGe detector in the planned application. For 662 keV photons, the spatial resolution obtained was approximately the same as the collimator slit width for both 100 and 150 mm long collimators. The corresponding spatial resolution at 1596 keV photon energy was almost twice the slit width for 100 mm collimator, due to the partial penetration of the high-energy gamma rays through the collimator bulk. For a 150 mm long collimator, an improved resolution was obtained.

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“…For accurate analysis of the source activity using gamma spectrometry, the self-attenuation needs to be accounted for when calculating the source activity from the measured peak intensity. In the case of tomography, it equivalently needs to be accounted for in the calculation of the system matrix (16). One method for this purpose is the analytical correction according to Eq.…”

Section: Theorymentioning

confidence: 99%

The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup

Atak

Anastasiadis

Jansson

et al. 2020

Progress in Nuclear Energy

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Nondestructive gamma-ray spectrometry of nuclear fuel is routinely performed in axial gamma scanning devices and more recently with gamma emission tomography. Following the irradiation of a fresh nuclear fuel with high intensity neutron flux in a nuclear reactor core, a great number of gamma-emitting radionuclides are created. These can be utilized for gamma spectrometric techniques. However, due to the high density and atomic number of the nuclear fuel, self-attenuation of gamma-rays is a challenge, which requires attenuation correction in order to perform accurate analysis of the source activity in the fuel.

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A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor

Abstract: A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor

Cited by 13 publications

References 6 publications

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

High-Fidelity Monte Carlo Analysis of Highly-Localized Fission Power Peaking near a Flux Trap in an MTR Core

Calculation of Spatial Response of a Collimated Segmented HPGe Detector for Gamma Emission Tomography by MCNP Simulations

The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup

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