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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.

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. Finansiering: Svenskt Kärntekniskt Centrum, SKC

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.

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.

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.

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.

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.