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Atak, H., Anastasiadis, A., Jansson, P., Elter, Z., Andersson Sundén, E., Holcombe, S. & Andersson, P. (2019). The degradation of gamma-ray mass attenuation of UO2 and MOX fuel with nuclear burnup.
Open this publication in new window or tab >>The degradation of gamma-ray mass attenuation of UO2 and MOX fuel with nuclear burnup
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2019 (English)In: Article in journal (Refereed) Submitted
National Category
Engineering and Technology Natural Sciences
Identifiers
urn:nbn:se:uu:diva-395145 (URN)
Note

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.

In this study, the degradation of the gamma-ray mass attenuation with burnup was investigated and, in addition, a predictive model was created by investigating the attenuation change at various gamma energies caused by the burnup of the nuclear fuel. This model is intended for use by spectrometry practitioners inspecting nuclear fuel. To this aim, the energy-dependent gamma-ray mass-attenuation coefficients were investigated as a function of burnup for UO2, and three MOX fuels having different initial Pu contents. The Serpent 2 reactor physics code was used to simulate the burnup history of the fuel pins. The nuclide inventory of the Serpent 2 output is combined with the NIST XCOM database to calculate the mass attenuation coefficients.

The mass attenuation coefficient of the fuel was found to decrease with the fuel burnup, in the range of a few percent, depending on the burnup and gamma energy. Also, a theoretical burnup dependent swelling model was imposed on fuel density to see how linear attenuation coefficient of fuel material is changed. Furthermore, greater effect may be expected on the transmitted intensity, where a simulation study of a PWR assembly revealed that the contribution from the inner rods in a scanned fuel assembly increased by tens of percent compared to the one with non-irradiated fresh fuels, when shielded by the outer rods of the assembly. A sensitivity analysis was performed in order to test the effect of a number of geometrical and operational reactor parameters that were considered to potentially effect the mass attenuation coefficient. Finally, a simple-to-use predictive model was constructed providing the mass-attenuation coefficient [cm2/g] of fuel as a function of burnup [MWd/kgHM] and initial Pu content [wt%].  The resulting predictive model was optimized by using the nonlinear regression method.

Available from: 2019-10-14 Created: 2019-10-14 Last updated: 2019-10-15Bibliographically approved
Andersson, P. (2019). Validation of axial flux profiles and development of a modified axial flux model using gamma scans of IFA-677. Institutt for energiteknikk
Open this publication in new window or tab >>Validation of axial flux profiles and development of a modified axial flux model using gamma scans of IFA-677
2019 (English)Report (Other academic)
Place, publisher, year, edition, pages
Institutt for energiteknikk, 2019
National Category
Engineering and Technology Physical Sciences
Identifiers
urn:nbn:se:uu:diva-374362 (URN)
Available from: 2019-01-21 Created: 2019-01-21 Last updated: 2019-01-21
Åberg Lindell, M., Andersson, P., Grape, S., Hellesen, C., Håkansson, A. & Eriksson, M. (2018). Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 885, 67-78
Open this publication in new window or tab >>Discrimination of irradiated MOX fuel from UOX fuel by multivariate statistical analysis of simulated activities of gamma-emitting isotopes
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2018 (English)In: Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, ISSN 0168-9002, E-ISSN 1872-9576, Vol. 885, p. 67-78Article in journal (Refereed) Published
Abstract [en]

This paper investigates how concentrations of certain fission products and their related gamma-ray emissions can be used to discriminate between uranium oxide (UOX) and mixed oxide (MOX) type fuel. Discrimination of irradiated MOX fuel from irradiated UOX fuel is important in nuclear facilities and for transport of nuclear fuel, for purposes of both criticality safety and nuclear safeguards. Although facility operators keep records on the identity and properties of each fuel, tools for nuclear safeguards inspectors that enable independent verification of the fuel are critical in the recovery of continuity of knowledge, should it be lost. A discrimination methodology for classification of UOX and MOX fuel, based on passive gamma-ray spectroscopy data and multivariate analysis methods, is presented. Nuclear fuels and their gamma-ray emissions were simulated in the Monte Carlo code Serpent, and the resulting data was used as input to train seven different multivariate classification techniques. The trained classifiers were subsequently implemented and evaluated with respect to their capabilities to correctly predict the classes of unknown fuel items. The best results concerning successful discrimination of UOX and MOX-fuel were acquired when using non-linear classification techniques, such as the k nearest neighbors method and the Gaussian kernel support vector machine. For fuel with cooling times up to 20 years, when it is considered that gamma-rays from the isotope  134Cs can still be efficiently measured, success rates of 100% were obtained. A sensitivity analysis indicated that these methods were also robust.

Keywords
Spent nuclear fuel, MOX, Gamma spectroscopy, Multivariate analysis, Classification
National Category
Physical Sciences
Identifiers
urn:nbn:se:uu:diva-337676 (URN)10.1016/j.nima.2017.12.020 (DOI)000424740800009 ()
Funder
Swedish Research Council, VR 621-2009-3991Swedish Radiation Safety Authority, SSM2016-661
Available from: 2018-01-03 Created: 2018-01-03 Last updated: 2018-04-19Bibliographically approved
Åberg Lindell, M., Andersson, P., Grape, S., Håkansson, A. & Eriksson, M. (2018). Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 897, 85-91
Open this publication in new window or tab >>Estimating irradiated nuclear fuel characteristics by nonlinear multivariate regression of simulated gamma-ray emissions
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2018 (English)In: Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, ISSN 0168-9002, E-ISSN 1872-9576, Vol. 897, p. 85-91Article in journal (Refereed) Published
Abstract [en]

In addition to verifying operator declared parameters of spent nuclear fuel, the ability to experimentally infer such parameters with a minimum of intrusiveness is of great interest and has been long-sought after in the nuclear safeguards community. It can also be anticipated that such ability would be of interest for quality assurance in e.g. recycling facilities in future Generation IV nuclear fuel cycles. One way to obtain information regarding spent nuclear fuel is to measure various gamma-ray intensities using high-resolution gamma-ray spectroscopy. While intensities from a few isotopes obtained from such measurements have traditionally been used pairwise, the approach in this work is to simultaneously analyze correlations between all available isotopes, using multivariate analysis techniques. Based on this approach, a methodology for inferring burnup, cooling time, and initial fissile content of PWR fuels using passive gamma-ray spectroscopy data has been investigated. PWR nuclear fuels, of UOX and MOX type, and their gamma-ray emissions, were simulated using the Monte Carlo code Serpent. Data comprising relative isotope activities was analyzed with decision trees and support vector machines, for predicting fuel parameters and their associated uncertainties. From this work it may be concluded that up to a cooling time of twenty years, the 95% prediction intervals of burnup, cooling time and initial fissile content could be inferred to within approximately 7 MWd/kgHM, 8 months, and 1.4 percentage points, respectively. An attempt aiming to estimate the plutonium content in spent UOX fuel, using the developed multivariate analysis model, is also presented. The results for Pu mass estimation are promising and call for further studies.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE BV, 2018
Keywords
Nuclear safeguards, Spent nuclear fuel, Gamma-ray, Multivariate analysis, Nonlinear regression
National Category
Subatomic Physics
Identifiers
urn:nbn:se:uu:diva-357374 (URN)10.1016/j.nima.2018.04.034 (DOI)000433206800014 ()
Available from: 2018-08-24 Created: 2018-08-24 Last updated: 2018-08-24Bibliographically approved
Andersson, P., Holcombe, S. & Tverberg, T. (2018). Inspection of LOCA Test Rod IFA-650.16 Using Gamma Emission Tomography.
Open this publication in new window or tab >>Inspection of LOCA Test Rod IFA-650.16 Using Gamma Emission Tomography
2018 (English)Report (Other academic)
National Category
Engineering and Technology
Identifiers
urn:nbn:se:uu:diva-383808 (URN)
Available from: 2019-05-22 Created: 2019-05-22 Last updated: 2019-05-22
Andersson, P. & Holcombe, S. (2017). A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor. Annals of Nuclear Energy, 110, 88-97
Open this publication in new window or tab >>A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor
2017 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 110, p. 88-97Article in journal (Refereed) Published
Abstract [en]

The Halden reactor project (HRP) has recently developed a gamma emission tomography instrument dedicated for measurements of irradiated nuclear fuel in collaboration with Westinghouse and Uppsala University. This instrument is now assembled and the first experimental measurements have been performed on fuel assemblies irradiated in the Halden reactor. The objective of the instrument is to map the distribution of radioisotopes of interest in the fuel, e.g. 137Cs or 140La/Ba, and for this purpose, a spectroscopic high-purity Germanium detector has been selected, which enables the identification and tomographic reconstruction of separate isotopes by their characteristic gamma rays.

To gain from the analysis of the data from this new instrument, and in the future from other gamma emission tomography instruments for nuclear fuels, various reconstruction methods are available that vary in the accuracy and the amount of detail obtainable in the analysis. This paper presents the details of the working principles of a new code for gamma emission tomography, the UPPREC (UPPsala university REConstruction) code. It is a development in MATLABTM code with the aim to produce detailed quantitative images of the investigated fuel.

In this paper, the methods assembled for the analysis of data collected by this novel instrument are described and demonstrated and a benchmark is presented using single rod gamma scanning. It is shown that the UPPREC methodology improves the accuracy of the reconstructions by removing the errors introduced by the presence of highly attenuating fuel and structural material in the fuel assembly. With the introduction of UPPREC, detailed quantitative cross-sectional images of nuclide concentrations in nuclear fuel are now able to be obtained by nondestructive means. This has potential applications in both nuclear fuel diagnostics and in safeguards.

National Category
Engineering and Technology
Identifiers
urn:nbn:se:uu:diva-313485 (URN)10.1016/j.anucene.2017.06.025 (DOI)000412251000010 ()
Note

Finansiering: Svenskt Kärntekniskt Centrum, SKC

Available from: 2017-01-20 Created: 2017-01-20 Last updated: 2019-10-18Bibliographically approved
Mattera, A., Pomp, S., Lantz, M., Rakopoulos, V., Solders, A., Al-Adili, A., . . . Eronen, T. (2017). A neutron source for IGISOL-JYFLTRAP: Design and characterisation. European Physical Journal A, 53(173)
Open this publication in new window or tab >>A neutron source for IGISOL-JYFLTRAP: Design and characterisation
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2017 (English)In: European Physical Journal A, ISSN 1434-6001, E-ISSN 1434-601X, Vol. 53, no 173Article in journal (Refereed) Published
Abstract [en]

A white neutron source based on the Be(p,nx) reaction for fission studies at the IGISOLJYFLTRAP facility has been designed and tested. 30 MeV protons impinge on a 5mm thick water-cooled beryllium disc. The source was designed to produce at least 1012 fast neutrons/s on a secondary fission target, in order to reach competitive production rates of fission products far from the valley of stability.

The Monte Carlo codes MCNPX and FLUKA were used in the design phase to simulate the neutron energy spectra. Two experiments to characterise the neutron field were performed: the first was carried out at The Svedberg Laboratory in Uppsala (SE), using an Extended-Range Bonner Sphere Spectrometer and a liquid scintillator which used the time-of-flight (TOF) method to determine the energy of the neutrons; the second employed Thin-Film Breakdown Counters for the measurement of the TOF, and activation foils, at the IGISOL facility in Jyväskylä (FI). Design considerations and the results of the two characterisation measurements are presented, providing benchmarks for the simulations.

National Category
Subatomic Physics
Identifiers
urn:nbn:se:uu:diva-328569 (URN)10.1140/epja/i2017-12362-x (DOI)000408661200001 ()
Funder
Swedish Radiation Safety AuthoritySwedish Nuclear Fuel and Waste Management Company, SKB
Available from: 2017-08-26 Created: 2017-08-26 Last updated: 2017-12-01Bibliographically approved
Andersson, P. & Holcombe, S. (2017). Feasibility Study of Using Gamma Emission Tomography for Identification of Leaking Fuel Rods in Commercial Fuel Assemblies. In: : . Paper presented at Water Reactor Fuel Performance Meeting (WRFPM), 10-14 September 2017, Jeju Island Korea.
Open this publication in new window or tab >>Feasibility Study of Using Gamma Emission Tomography for Identification of Leaking Fuel Rods in Commercial Fuel Assemblies
2017 (English)Conference paper, Published paper (Refereed)
National Category
Other Engineering and Technologies not elsewhere specified
Identifiers
urn:nbn:se:uu:diva-328197 (URN)
Conference
Water Reactor Fuel Performance Meeting (WRFPM), 10-14 September 2017, Jeju Island Korea
Available from: 2017-08-18 Created: 2017-08-18 Last updated: 2017-09-14Bibliographically approved
Andersson, P. (2017). Gamma Emission Tomography of LOCA-transient test rods. In: : . Paper presented at Nationell strålsäkerhet : utblick och forskning, 22–23 november 2017, Stockholm.
Open this publication in new window or tab >>Gamma Emission Tomography of LOCA-transient test rods
2017 (English)Conference paper, Oral presentation only (Other (popular science, discussion, etc.))
National Category
Accelerator Physics and Instrumentation
Identifiers
urn:nbn:se:uu:diva-338666 (URN)
Conference
Nationell strålsäkerhet : utblick och forskning, 22–23 november 2017, Stockholm
Available from: 2018-01-11 Created: 2018-01-11 Last updated: 2018-01-17Bibliographically approved
Andersson, P. (2017). Notes on the world energy supply.
Open this publication in new window or tab >>Notes on the world energy supply
2017 (English)Other (Other (popular science, discussion, etc.))
National Category
Engineering and Technology
Identifiers
urn:nbn:se:uu:diva-326287 (URN)
Available from: 2017-07-05 Created: 2017-07-05 Last updated: 2017-07-11Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0001-7370-6539

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