| 1. |
- Kanai, M, et al.
(författare)
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- 2023
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swepub:Mat__t
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| 2. |
- Niemi, MEK, et al.
(författare)
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- 2021
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swepub:Mat__t
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| 3. |
- Andersson, Peter, 1981-, et al.
(författare)
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A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor
- 2017
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 88-97
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 4. |
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| 5. |
- Andersson, Peter, 1981-, et al.
(författare)
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Inspection of a LOCA Test Rod at the Halden Reactor Project using Gamma Emission Tomography
- 2016
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Ingår i: Top Fuel 2016 - LWR Fuels with Enhanced Safety and Performance. - : American Nuclear Society. - 9780894487347
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Konferensbidrag (refereegranskat)abstract
- The LOCA test series IFA-650 conducted at the OECD Halden Reactor Project (HRP) has provided unique data on the performance of fuel rods during LOCA transients. One focus of the current investigations is the performance of the fuel in the ballooning stage of the LOCA transient. In this stage, relocation of fuel material is a possibility, in which case pellet fragments fall down to fill the void introduced by the increased volume of the ballooned cladding. This increases the heat load in that region, further promoting corrosion of the cladding. A special concern in the case of high-burnup fuels is the increasing number of small fuel fragments, which may be expected to cause a higher packing fraction in the ballooned region. In this work, a novel technique is presented for assessing the average density of the fuel material in the ballooned region of LOCA test rods. The investigation is based on non-destructive gamma emission tomography measurements, using the dedicated instrument recently developed at the HRP in collaboration with Westinghouse (Sweden) and Uppsala University. In this approach, the gamma radiation field surrounding the test rod has been measured with a narrowly collimated HPGe detector. Tomographic reconstruction of the data was performed, providing the radial gamma-ray source distribution within the measured volume, which reveals the fuel fragment distribution. From this, the density of the fuel in the measured volume (i.e., the packing fraction) may be calculated. The technique has been used to investigate a LOCA test rod of the Halden Reactor Project LOCA series. The LOCA experiment was carried out about one month prior to the gamma tomography examination. The results show that the distribution of the relocated fuel can be imaged using gamma rays from fission products. The reconstructions of the 662 keV rays from 137Cs and 1596 keV from 140Ba/La are demonstrated. In addition, the peaks of activation products offer valuable information on the location of the test rig structures, which may be utilized in a quantitative tomographic reconstruction to assess the spatially resolved packing fraction.
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| 6. |
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| 7. |
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| 8. |
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| 9. |
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| 10. |
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| 11. |
- Andersson, Peter, 1981-, et al.
(författare)
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Simulation of the response of a segmented High-Purity Germanium detector for gamma emission tomography of nuclear fuel
- 2020
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Ingår i: SN Applied Sciences. - : Springer. - 2523-3963 .- 2523-3971. ; 2
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Tidskriftsartikel (refereegranskat)abstract
- Irradiation testing of nuclear fuel is routinely performed in nuclear test reactors. For qualification and licensing of Accident Tolerant Fuels or Generation IV reactor fuels, an extensive increase in irradiation testing is foreseen in order to fill the gaps of existing validation data, both in normal operational conditions and in order to identify operational limits.Gamma Emission Tomography (GET) has been demonstrated as a viable technique for studies of the behavior of irradiated nuclear fuel, e.g. measurement of fission gas release and inspection of fuel behavior under Loss-Of-Coolant Accident conditions. In this work, the aim is to improve the technique of GET for irradiated nuclear fuel by developing a detector concept for an improved tomography system that allows for a higher spatial resolution and/or faster interrogation.We present the working principles of a novel concept for gamma emission tomography using a segmented High Purity Germanium (HPGe) detector. The performance of this concept was investigated using the Monte Carlo particle transport code MCNP. In particular, the data analysis of the proposed detector was evaluated, and the performance, in terms of full energy efficiency and localization failure rate, has been evaluated.We concluded that the segmented HPGe detector has an advantageous performance as compared to the traditional single-channel detector systems. Due to the scattering nature of gamma rays, a trade-off is presented between efficiency and cross-talk; however, the performance is nevertheless a substantial improvement over the currently used single-channel HPGe detector systems.
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| 12. |
- Atak, Haluk, et al.
(författare)
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The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup
- 2020
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Ingår i: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 125
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Tidskriftsartikel (refereegranskat)abstract
- 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 UOX, 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.
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| 13. |
- Davour, Anna, et al.
(författare)
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Applying image analysis techniques to tomographic images of irradiated nuclear fuel assemblies
- 2016
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 223-229
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Tidskriftsartikel (refereegranskat)abstract
- In this paper we present a set of image analysis techniques used for extraction of information from cross-sectional images of nuclear fuel assemblies, achieved from gamma emission tomography measurements. These techniques are based on template matching, an established method for identifying objects with known properties in images.We demonstrate a rod template matching algorithm for identification and counting of the fuel rods present in the image. This technique may be applicable in nuclear safeguards inspections, because of the potential of verifying the presence of all fuel rods, or potentially discovering any that are missing.We also demonstrate the accurate determination of the position of a fuel assembly, or parts of the assembly, within the imaged area. Accurate knowledge of the assembly position enables detailed modelling of the gamma transport through the fuel, which in turn is needed to make tomographic reconstructions quantifying the activity in each fuel rod with high precision.Using the full gamma energy spectrum, details about the location of different gamma-emitting isotopes within the fuel assembly can be extracted. We also demonstrate the capability to determine the position of supporting parts of the nuclear fuel assembly through their attenuating effect on the gamma rays emitted from the fuel. Altogether this enhances the capabilities of non-destructive nuclear fuel characterization.
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| 14. |
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| 15. |
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| 16. |
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| 17. |
- Holcombe, Scott, et al.
(författare)
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A Novel gamma emission tomography instrument for enhanced fuel characterization capabilities within the OECD Halden Reactor Project
- 2015
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 837-845
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Tidskriftsartikel (refereegranskat)abstract
- Gamma emission tomography is a method based on gamma-ray spectroscopy and tomographic reconstruction techniques, which can be used for rod-wise characterization of nuclear fuel assemblies without dismantling the fuel. By performing a large number of measurements of the gamma-ray flux intensity around a fuel assembly using a well-collimated gamma-ray detector, the internal source distribution in the assembly may be reconstructed using tomographic algorithms. If a spectroscopic detection system is used, different gamma-ray emitting isotopes can be selected for analysis, enabling nondestructive fuel characterization with respect to a variety of fuel parameters. In this paper, we describe a novel gamma emission tomography instrument, which has been designed, constructed and tested at the Halden Boiling Water Reactor (HBWR). The device will be used to characterize fuel assemblies irradiated in the HBWR as part of ongoing nuclear fuel research conducted within the OECD Halden Reactor Project (HRP). As compared to single-rod gamma scanning, where the fuel is dismantled and the gamma radiation from each rod is measured separately, handling time associated with characterizing the fuel can be significantly reduced when using the gamma emission tomography device. Furthermore, because gamma emission tomography enables rod-wise fuel characterization without dismantling, even instrumented experimental fuel assemblies may be characterized repeatedly throughout the fuel's lifetime, with limited risk of damaging the fuel or its instrumentation. Accordingly, the capabilities of fuel characterization within the OECD HRP are expected to be strongly enhanced by the deployment of this device. Here, the gamma-tomographic method and the experimental setup are demonstrated through experimental measurements of the fuel stack and gas plenum regions of a nine-rod HBWR fuel assembly configuration, where four rods had a burnup of approximately 26 MWd/kgUO(2) and five rods had a burnup of approximately 50 MWd/kgUO(2). Tomographic images are presented, which show the applicability for assessment of fission gas contents in the gas plena and of fission products in the fuel stack. Furthermore, neutron activation products are analyzed, which give additional information on construction material properties.
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| 18. |
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| 19. |
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| 20. |
- Holcombe, Scott, et al.
(författare)
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Determination of the Rod-wise Fission Gas Release Fraction in a Complete Fuel Assembly Using Non-destructive Gamma Emission Tomography
- 2016
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 837, s. 99-108
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Tidskriftsartikel (refereegranskat)abstract
- A gamma tomography instrument has been developed at the Halden Boiling Water Reactor (HBWR) in cooperation between the Institute for Energy Technology, Westinghouse (Sweden) and Uppsala University. The instrument is used to record the gamma radiation field surrounding complete fuel assemblies and consists of a shielded enclosure with fixtures to accurately position the fuel and detector relative to each other. A High Purity Germanium detector is used for acquiring high-resolution spectroscopic data, allowing for analysis of multiple gamma-ray peaks. Using the data extracted from the selected peaks, tomographic reconstruction algorithms are used to reproduce the corresponding spatial gamma-ray source distributions within the fuel assembly. With this method, rod-wise data can be can be deduced without the need to dismantle the fuel.In this work, the tomographic device has been experimentally benchmarked for non-destructive rod-wise determination of the Fission Gas Release (FGR) fraction. Measurements were performed on the fuel-stack and gas-plenum regions of a complete fuel assembly, and quantitative tomographic reconstructions of the measurement data were performed in order to determine the rod-wise ratio of 85Kr in the gas plenum to 137Cs in the fuel stack. The rod-wise ratio of 85Kr/137Cs was, in turn, used to calculate the rod-wise FGR fraction. In connection to the tomographic measurements, the fuel rods were also measured individually using gamma scanning in order to provide an experimental benchmark for the tomographic method.Fuel rods from two donor driver fuel assemblies were placed into a nine-rod HBWR driver fuel assembly configuration. In order to provide a challenging measurement object and thus an appropriate benchmark for the tomographic method, five rods were taken from an assembly with a burnup of 51 MWd/kgUO2, and four rods were from an assembly with a burnup of 26 MWd/kgUO2. At the time of the measurements, the nine rods had cooled for approximately 22 years. All fuel rods had operated at high linear heat rates (around 70 kW/m), thus leading to relatively high FGR fractions. Here, the FGR fraction was determined to be ~24% in the high-burnup rods, and ~17% in the low-burnup rods. The tomography measurement results were in good agreement with the results from individual rod scanning, demonstrating the feasibility of tomography for this application. The capability of tomography to assess individual fuel rods without the need to dismantle the assembly can be particularly valuable in cases of fuels that do not allow disassembly, such as experimental HBWR fuel fitted with extensive instrumentation.
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| 21. |
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| 22. |
- Holcombe, Scott, et al.
(författare)
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Feasibility of identifying leaking fuel rods using gamma tomography
- 2013
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 334-340
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Tidskriftsartikel (refereegranskat)abstract
- In cases of fuel failure in irradiated nuclear fuel assemblies, causing leakage of fission gasses from a fuel rod, there is a need for reliable non-destructive measurement methods that can determine which rod is failed. Methods currently in use include visual inspection, eddy current, and ultrasonic testing, but additional alternatives have been under consideration, including tomographic gamma measurements.The simulations covered in this report show that tomographic measurements could be feasible. By measuring a characteristic gamma energy from fission gasses in the gas plenum, the rod-by-rod gamma source distribution within the fuel rod plena may be reconstructed into an image or data set which could then be compared to the predicted distribution of fission gasses, e.g. from the STAV code. Rods with significantly less fission gas in the plenum may then be identified as leakers.Results for rods with low fission gas release may, however, in some cases be inconclusive since these rods will already have a weak contribution to the measured gamma-ray intensities and for such rods there is a risk that a further decrease in fission gas content due to a leak may not be detectable. In order to evaluate this and similar experimental issues, measurement campaigns are planned using a tomographic measurement system at the Halden Boiling Water Reactor.
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| 23. |
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| 24. |
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| 25. |
- Holcombe, Scott, et al.
(författare)
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Gamma Emission Tomography Measurements of Fuel Assemblies at the Halden Reactor
- 2016
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Ingår i: LWR Fuels with Enhanced Safety and Performance Meeting (Top Fuel 2016). - La Grange Park, Illinois : American Nuclear Society (ANS). - 9781510842168 ; , s. 1601-1611
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Gamma emission tomography measurements have been performed at the Halden Reactor Project using a dedicated instrument, which utilizes a high-resolution gamma-ray detector allowing for spectroscopic analysis of the measurement data. Measurements have been performed on two fuel assemblies consisting of 9 and 13 rods, respectively, in order to characterize the rod-wise radioactive source distribution (i.e. fission and activation products) within the fuel. The 9-rod fuel assembly consisted of five fuel rods at 50 MWd/kgUO2 and four fuel rods at 26 MWd/kgUO2, where all rods had cooled for approximately 22 years at the time of measurement. The rods in the 13-rod assembly all had a burnup of ~4.5 MWd/kgUO2, and cooled for approximately 1.5 years at the time of measurement.In this paper, the tomographic data was reconstructed using the Filtered Backprojection technique, where no consideration to gamma-ray attenuation in the fuel was given. Due to the varying burnup and cooling times between the assemblies, the spectroscopic data also varied between the respective sets of measurements. The high-resolution detector used in the measurements allowed for tomographic reconstruction of many gamma-ray peaks corresponding to various fission products and activation products present in the fuel and structural materials.The qualitative tomographic images presented in this paper are analyzed to determine the positions of the fuel rods and structural components in the fuel. This geometrical information will subsequently be used as input to algebraic reconstruction algorithms which are used to determine the quantitative rod-wise gamma-ray source distributions.The gamma tomography instrument in Halden was designed, constructed, and demonstrated in collaboration between the Westinghouse (Sweden), and Uppsala University.
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| 26. |
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| 27. |
- Holcombe, Scott
(författare)
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Gamma Spectroscopy and Gamma Emission Tomography for Fuel Performance Characterization of Irradiated Nuclear Fuel Assemblies
- 2014
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Gamma spectroscopy and gamma emission tomography are two non-destructive measurement techniques for assessing the performance of nuclear fuel which have been investigated in this thesis for existing and novel applications through theoretical studies and experimental demonstrations. For assessment of individual fuel rods using gamma spectroscopy, fuel assemblies are dismantled so that the fuel rods may be measured separately, which is time-consuming and may cause damage to the fuel. Gamma tomography is more seldom used, but its application on complete fuel assemblies would enable the assessment of individual fuel rods without the need to disassemble the fuel. Both techniques are based on recording gamma rays, emitted at characteristic energies from decaying radioactive products in the fuel.The feasibility of measuring short-lived fission gasses in the gas plenum of fuel rods with short cooling time was experimentally investigated. Based on the feasibility demonstration, a method was proposed and experimentally demonstrated for determining the fission gas release fraction of 133Xe in fuel rods with short cooling time. Additionally, a method for investigating the origin of released fission gasses based on the measured ratio of 133Xe/85Kr in the fuel rod gas plenum was demonstrated. These methods may be employed at research reactors, where fuel with short cooling time is available for measurement.A gamma emission tomography instrument has been designed, constructed and experimentally demonstrated on a Halden Reactor fuel assembly. Simulation studies showed that the instrument and the tomographic reconstruction methods employed may be useful for: identifying a leaking fuel rod in an assembly by its lack of fission gas content; reconstruction of the rod-wise fission product distributions in the fuel stack and plenum regions of the assembly; and determining the rod-wise fission gas release fractions.In the experimental demonstration, the rod-wise distributions of the fission products 137Cs and 85Kr in the fuel stack and plenum regions of the assembly were reconstructed, as well as the distributions of the activation products 60Co and 178mHf in the plenum region, revealing the plenum springs and tie rods, respectively. The reconstructed data was in the form of images, useful for qualitative assessment of the fuel.
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| 28. |
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| 29. |
- Holcombe, Scott, et al.
(författare)
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Method For Analyzing Fission Gas Release In Fuel Rods Based On Gamma-Ray Measurements Of Short-Lived Fission Products
- 2013
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 184:1, s. 96-106
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Tidskriftsartikel (refereegranskat)abstract
- Fission gases are produced as a result of fission reactions in nuclear fuel. Most of these gases remain trapped within the fuel pellets, but some may be released to the fuel rod internal gas volume under certain conditions. This phenomenon of fission gas release is important for fuel performance since the released gases can degrade the thennal properties of the fuel rod. fill gas and contribute to increasing fuel rod internal pressure. Various destructive and nondestructive methods are available for determining the amount of fission gas release; however, the current methods are primarily useful for determining the integrated fission gas release fraction, i.e., the amount of fission gas produced in the fuel that has been released to the free rod volume over the entire lifetime of a nuclear fuel rod. In this work, a method is proposed for determining the fission gas release that occurs during short irradia-tion sequences. The proposed method is based on spectroscopic measurements of gamma rays emitted in the decay of short-lived fission gas isotopes. Determining such sequence-specific fission gas release can be of interest when evaluating the fuel behavior for selected times during irradiation, such as during power ramps. The data obtained in this type of measurement may also be useful for investigating the mechanisms behind fission gas release for fuel at high burnup. The method is demonstrated based on the analysis of experimental gamma-ray spectra previously collected using equipment not dedicated for this purpose; however, the analysis indicates the feasibility of the method. Further evaluation of the method is planned, using dedicated equipment at the Halden Boiling Water Reactor.
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| 30. |
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| 31. |
- Holcombe, Scott
(författare)
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Nondestructive Fission Gas Measurements by Means of Gamma Spectroscopy and Gamma Tomography
- 2012
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- There is a continuous need in the nuclear industry to characterize irradiated nuclear fuel rods and assemblies, both for fuel performance and for safeguards purposes, and consequently there are various destructive and nondestructive measurement techniques available to meet this need. Gamma spectroscopy is one such nondestructive technique, which has been extensively used for a variety of fuel characterization applications. Furthermore, gamma tomography – a combination of gamma spectroscopic measurements and tomographic reconstruction – has in recent years been demonstrated as an efficient technique for characterization of irradiated nuclear fuel assemblies on a rod-by-rod basis without the need to dismantle the fuel. This thesis comprises four scientific papers in which novel applications of these two techniques have been developed and evaluated.The major part of this work has been performed at the Halden Boiling Water Reactor (HBWR), where a gamma tomography measurement system is currently under construction, as presented in this thesis. The methods and evaluations presented in this work are based on the conditions at the HBWR.Based on gamma spectroscopy, a novel nondestructive method for determining fission gas release which occurs over short irradiation sequences has been developed, comprising the measurement and analysis of short lived isotopes in individual fuel rods. The method has been demonstrated based on gamma-ray spectra recorded from an experimental fuel rod irradiated in the HBWR.Based on gamma tomography, a novel method for identifying failed fuel rods within a nuclear fuel assembly has also been developed. The method comprises the measurement of gamma rays emitted in the decay of selected fission gas isotopes in the gas plenum region of a fuel assembly, tomographic image reconstruction of the internal source distribution and subsequent analysis of the resulting image in order to determine if any of the fuel rods in the assembly has unexpectedly low activity, indicating that it is a leaking fuel rod. Simulation studies performed for HBWR fuel show highly promising results for gamma rays emitted in the decay of two selected fission gas isotopes.The methods will be further investigated at the HBWR, by performing dedicated gamma spectroscopy measurements and by using the tomographic measurement system currently under construction.
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| 32. |
- Insulander Björk, Klara L, 1982, et al.
(författare)
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Irradiation testing of enhanced uranium oxide fuels
- 2019
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 125, s. 99-106
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Tidskriftsartikel (refereegranskat)abstract
- Enhanced uranium oxide fuel types are being tested in the Halden Research Reactor in Norway with the aim is to assess the effect that these enhancements have on fuel performance. Fuel temperatures, rod pressures and dimensional changes are being monitored online and an extensive post-irradiation examination programme is planned. Preliminary data show that fuel centerline temperatures can be lowered by addition of ThO2 to the fuel matrix, or by incorporating Cr or SiO2-TiO2 as a network structure within the fuel. In parallel, two types of cladding coatings are tested in order to investigate their in-core properties. No abnormal behaviour has been noted during the first 100 days of irradiation.
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| 33. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies
- 2015
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier. - 0168-9002 .- 1872-9576. ; 783, s. 128-141
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Tidskriftsartikel (refereegranskat)abstract
- A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly׳s internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for.As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel.Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies׳ completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which may be particularly useful in the latter application.Two main classes of algorithms are covered; (1) analytic filtered back-projection algorithms, and (2) a group of model-based or algebraic algorithms. For the former class, a basic algorithm has been implemented, which does not take attenuation in the materials of the fuel assemblies into account and which assumes an idealized imaging geometry. In addition, a novel methodology has been presented for introducing a first-order correction to the obtained images for these deficits; in particular, the effects of attenuation are taken into account by modelling the response for an object with a homogeneous mix of fuel materials in the image area. Neither the basic algorithm, nor the correction method requires prior knowledge of the fuel geometry, but they result in images of the assembly׳s internal activity distribution. Image analysis is then applied to deduce quantitative information.Two algebraic algorithms are also presented, which model attenuation in the fuel assemblies to different degrees; either assuming a homogenous mix of materials in the image area without a priori information or utilizing known information of the assembly geometry and of its position in the measuring setup for modelling the gamma-ray attenuation in detail. Both algorithms model the detection system in detail. The former algorithm returns an image of the cross-section of the object, from which quantitative information is extracted, whereas the latter returns conclusive relative rod-by-rod data.Here, all reconstruction methods are demonstrated on simulated data of a 96-rod fuel assembly in a tomographic measurement setup. The assembly was simulated with the same activity content in all rods for evaluation purposes. Based on the results, it is argued that the choice of algorithm to a large degree depends on application, and also that a combination of reconstruction methods may be useful. A discussion on alternative analysis methods is also included.
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| 34. |
- Jacobsson Svärd, Staffan, 1972-, et al.
(författare)
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Gamma-ray Emission Tomography: Modelling and evaluation of partial-defect testing capabilities
- 2014
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Assessment of gamma emission tomography (GET) for spent nuclear fuel verification is the task in IAEA MSP project JNT1955. In line with IAEA Safeguards R&D plan 2012-2023, the aim of this effort is to “develop more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect tests on spent fuel assemblies prior to transfer to difficult to access storage". The current viability study constitutes the first phase of three, with evaluation and decision points between each phase. Two verification objectives have been identified; (1) counting of fuel pins in tomographic images without any a priori knowledge of the fuel assembly under study, and (2) quantitative measurements of pin-by-pin properties, e.g. burnup, for the detection of anomalies and/or verification of operator-declared data.Previous measurements performed in Sweden and Finland have proven GET highly promising for detecting removed or substituted fuel pins (i.e. partial defects) in BWR and VVER-440 fuel assemblies even down to the individual fuel pin level. The current project adds to previous experiences by pursuing a quantitative assessment of the capabilities of GET for partial defect detection, across a broad range of potential IAEA applications, fuel types, and fuel parameters. A modelling and performance-evaluation framework has been developed to provide quantitative GET performance predictions, incorporating burn-up and cooling-time calculations, Monte Carlo radiation-transport and detector-response modelling, GET instrument definitions (existing and notional) and tomographic reconstruction algorithms, which use recorded gamma-ray intensities to produce cross-sectional images of the source distribution in the fuel assembly or conclusive pin-by-pin data. The framework also comprises image-processing algorithms and performance metrics that recognize the inherent trade-off between the probability of detecting missing pins and the false-alarm rate. Here, the modelling and analysis framework is described and preliminary results are presented.
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| 35. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Tomographic determination of spent fuel assembly pin-wise burnup and cooling time for detection of anomalies
- 2015
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The IAEA has initiated Member States’ Support Program project JNT A 1955 to assess the partial defect detection capabilities of gamma emission tomography (GET) for spent nuclear fuel assembly verification. The GET technique is based on measurements of the gamma-ray flux distribution around a spent fuel assembly using dedicated, tomographic equipment and subsequent reconstruction of the internal source distribution using tomographic algorithms applied on the recorded data. One of the verification objectives identified for the project is the quantitative measurement of pin-by-pin properties, e.g. burnup and/or cooling time, for the detection of anomalies and/or verification of operator-declared data. For this objective, reconstruction algorithms that return quantitative, isotopic pin-by-pin data are applied.Previously, GET measurements performed on commercial nuclear fuel assemblies in Sweden have proven capable of determining the relative pin-by-pin power distribution with high precision in BWR fuel with short cooling time, based on the measured distribution of the gamma-ray emitting fission product 140Ba/La in the fuel. In the current project, the capabilities of GET to determine additional pin-wise fuel parameters in additional fuel types are being assessed. The evaluations are based on Monte Carlo simulations of the emission of gamma-rays from the fuel and their detection in a tomographic measurement device.This paper describes the algorithms used for reconstructing quantitative pin-wise data and the results that are anticipated with this technique. It is argued that detailed modelling of the gamma-ray attenuation through the highly inhomogeneous mix of strongly-attenuating fuel rods and less-attenuating surrounding water (wet storage) or air (dry storage) is required to yield high precision in the reconstructed data. The burnup distribution assessment would be based on the recording of 662-keV gamma radiation from 137Cs, whereas the assessment of both burnup and cooling time simultaneously requires the GET measurement and pin-wise reconstruction of at least two isotopes, which puts constraints on the measurement equipment used.
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| 38. |
- Senis, Lorenzo, et al.
(författare)
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Evaluation of gamma-ray transmission through rectangular collimator slits for application in nuclear fuel spectrometry
- 2021
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier. - 0168-9002 .- 1872-9576. ; 1014
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Tidskriftsartikel (refereegranskat)abstract
- Gamma-ray spectrometry is widely applied in several science fields, and in particular in non-destructive gamma scanning and gamma emission tomography of irradiated nuclear fuel. Often, a collimator is used in the experimental setup, to selectively interrogate a region of interest in the fuel. For the optimization of instrument design, as well as for planning measurement campaigns, predictive models for the transmitted gamma-ray intensity through the collimator are needed. Commonly, Monte Carlo Radiation Transport tools are used for accurate prediction of gamma-ray transport, however, the long computation time requirements when used in low-efficiency experimental setups present challenges.In this work, the full-energy peak intensity transmitted through a rectangular collimator slit was examined. A uniform planar surface source emitting isotropically was considered, and the rate of photons reaching an ideal counter plane on the opposite side of the collimator was evaluated by analytical integration. To find a closed-form primitive function, some idealizations were required, and thereby parametric models were obtained for the optical field of view, dependent on slit dimensions (length, height and width) and source-to-collimator distance. It was shown that the count rate in the detector is independent of the collimator-to-source distance. For contributions from outside the optical field of view, where a closed-form expression cannot be found, instead fast numerical integration methods were proposed.The results were validated using the Monte Carlo code MCNP6. For the analytical method, deviations were larger, the shorter the collimator, with up to 25% of underestimation obtained for the shortest examined collimator of 10 cm length. However, the longer the collimator, the better the observed agreement. This accuracy is deemed to be sufficient for instrument design and measurement planning, where often the order of magnitude of the count rate is not a priori known. For the numerical method, the results showed an agreement within 3 % for all evaluated collimator settings. The methods are planned for use in iterative optimization routines in the design of Gamma Emission Tomography devices, as well as for the prediction of gamma spectra obtained in the planning of fuel inspections. An application of the proposed method was demonstrated in spectrum prediction for a short cooling-time fuel rod test from the Halden reactor.
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| 39. |
- Smith, Eric L., et al.
(författare)
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A Viability Study of Gamma Emission Tomography for Spent Fuel Verification : JNT 1955 Phase I Technical Report
- 2016
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Rapport (refereegranskat)abstract
- The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information about the assembly. The second objective is to provide quantitative assay of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives has been evaluated across a range of fuel types, burnups, and cooling times, and with a target total interrogation time of less than 60 minutes. This evaluation of GET viability for safeguards applications was founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. Instrument response data were processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives were used to predict performance. This report describes the analysis framework and metrics used to predict tomographer performance, the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA, and a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.
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