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- Holcombe, S., et al.
(författare)
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Advanced fuel assembly characterization capabilities based on gamma tomography at the halden boiling water reactor
- 2012
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Ingår i: Proc. Int. Conf. on Advances in Reactor Physics. - 9781622763894 ; , s. 3478-3489
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Konferensbidrag (refereegranskat)abstract
- Characterization of individual fuel rods using gamma spectroscopy is a standard part of the Post Irradiation Examinations performed on experimental fuel at the Halden Boiling Water Reactor. However, due to handling and radiological safety concerns, these measurements are presently carried out only at the end of life of the fuel, and not earlier than several days or weeks after its removal from the reactor core. In order to enhance the fuel characterization capabilities at the Halden facilities, a gamma tomography measurement system is now being constructed, capable of characterizing fuel assemblies on a rod-by-rod basis in a more timely and efficient manner. Gamma tomography for measuring nuclear fuel is based on gamma spectroscopy measurements and tomographic reconstruction techniques. The technique, previously demonstrated on irradiated commercial fuel assemblies, is capable of determining rod-by-rod information without the need to dismantle the fuel. The new gamma tomography system will be stationed close to the Halden reactor in order to limit the need for fuel transport, and it will significantly reduce the time required to perform fuel characterization measurements. Furthermore, it will allow rod-by-rod fuel characterization to occur between irradiation cycles, thus allowing for measurement of experimental fuel repeatedly during its irradiation lifetime. The development of the gamma tomography measurement system is a joint project between the Institute for Energy Technology - OECD Halden Reactor Project, Westinghouse (Sweden), and Uppsala University.
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| 4. |
- 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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| 6. |
- Jammes, C., et al.
(författare)
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Progress in the development of the neutron flux monitoring system of the French GEN-IV SFR: Simulations and experimental validations
- 2015
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Ingår i: 2015 4th International Conference on Advancements in Nuclear Instrumentation Measurement Methods and their Applications, ANIMMA 2015. - 9781479999187
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Konferensbidrag (refereegranskat)abstract
- The neutron flux monitoring system of the French GEN-IV sodium-cooled fast reactor will rely on high-temperature fission chambers installed in the reactor vessel and capable of operating over a wide-range neutron flux. The definition of such a system is presented and the technological solutions are justified with the use of simulation and experimental results. © 2015 IEEE.
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