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 MCNP6.2. 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.
The combined third and fourth irradiation in the Advanced Gas Reactor (AGR) program (AGR-3/4) contained tristructural isotropic (TRISO)-coated particle fuel and designed-to-fail (DTF) fuel particles. The DTF particles were only coated with highly anisotropic pyrocarbon (PyC) so they would purposely fail during the AGR-3/4 irradiation and provide a source of fission products for measurement. Frickey performed materialographic preparation of the irradiated and unirradiated fuel compacts. This included designing and fabricating the mounts, grinding, and polishing the mounts at the Hot Fuel Examination Facility at Idaho National Laboratory. Cassie Anderson-Thueson assisted in the mounting, grinding, polishing, and optical microscopy work. John Stanek and Francine Rice assisted in the mount design and provided valuable input to sample preparation and microscopy processes leading to acquisition of the images included in this report. Fuel burnup and fast neutron fluence were calculated by Dr. Jim Sterbentz. Fuel temperatures and temperature variations were calculated by Grant Hawkes.
The second Advanced Gas Reactor irradiation (AGR-2) featured UCO tristructural isotropic (TRISO)-coated particle fuel and, for comparison purposes, UO 2 TRISO fuel. Particles from three UCO fuel compacts and one UO 2 fuel compact were chosen for analysis of kernel swelling and buffer densification (also called buffer shrinkage). Both kernel swelling and buffer shrinkage are common elements of TRISO fuel performance models. More recently, the AGR program has determined that buffer shrinkage may impact the integrity of the inner pyrolytic carbon (IPyC) layer, which may impact SiC layer integrity (particularly during high-temperature heating tests). Therefore, measurements of buffer irradiation-induced shrinkage may be useful in improving existing models and developing new models that capture buffer-IPyC-SiC-fission product interactions observed in recent AGR post-irradiation examination.TRISO particles were mounted in epoxy, ground, polished, and imaged via optical microscopy in four separate iterations. Each of the four iterations revealed progressively deeper cross sections within the particles. Images collected from each iteration were analyzed for the circumferences/radii of the kernel, buffer, IPyC, and SiC. Spheres were fit to each set of measurements to generate spherical radii representative of the components of the fuel particle. From these radii, volumes were computed and compared to the as-fabricated volumes.UCO kernel swelling was similar among the three AGR-2 compacts ranging from approximately 28 ± 7% to 32 ± 9%. This is only slightly higher swelling than was measured in an earlier study of an AGR-1 compact. However, the burnup at which this swelling occurred in AGR-2 was lower than in AGR-1, indicating a faster swelling rate for the AGR-2 kernels over the irradiation conditions covered by these compacts. The average AGR-2 UO 2 kernel swelling was less, at 10 ± 10%, a mean value comparable to other studies of UO 2 kernel swelling.While AGR-2 UCO and UO 2 kernels demonstrated different rates of swelling, the extent of buffer shrinkage was similar among the two fuel types, ranging from 24 ± 26% to 28 ±20%. This is significantly less shrinkage than was observed for AGR-1 buffers. A commonly-observed post-irradiation particle morphology is one where the buffer pulls away from the IPyC (at least in some regions of the particle). Besides the possibility of the buffer degrading the IPyC when it pulls away, this gap is significant because it may also reduce heat transfer within the particle or alter transport of fission products from the buffer to the IPyC. Average AGR-2 buffer-IPyC gaps were similar for the UCO and UO 2 particles, ranging from 22 ± 6 μm to 26 ± 4 μm. This is similar to what was measured for AGR-1 Compact 1-3-1. vii ACKNOWLEDGEMENTSCritical contributions to this work were made by a number of people. Particles for this work were shipped from Oak Ridge National Laboratory (where Dr. John Hunn was the primary point of contact) to INL. Brian Frickey performed the materiolographic mounting (includ...
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