| 1. |
- Wolniewicz, Peter, 1978-, et al.
(författare)
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Reactivity changes in lead-cooled fast reactors due to bubbles in the coolant
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Annan publikation (övrigt vetenskapligt/konstnärligt)abstract
- The formation of bubbles in the coolant of a Lead-Cooled Fast Reactor (LFR) may originate from a leaking heat-exchanger and is a potential safety hazard. Small bubbles can travel with the coolant without escaping to the cover gas, causing an increasing effective voiding of the coolant in a homogeneous manner. If the small bubbles coalesce into a larger bubble located at a stagnation zone, the reactor core may eventually be exposed to a transient bubble travelling axially through the core with a resulting change in the reactivity of the system. This study is focused on the reactivity changes caused by bubbles of various sizes and for different vertical positions as the bubble rises through the core. Three different sizes of LFR’s; 50 MWth, 300 MWth and 1200 MWth,respectively were user for the study. The 300 MWth reactor design is based on the Advanced LFR European Demonstrator (ALFRED) and the two other reactors are scaled up and scaled down versions of it and these were simulated in order study the sensitivity to void as a function of reactor size. We show that LFR’s may have a positive reactivity response to transient bubbles and that the sensitivity to changes in reactivity is larger the smaller the reactor. For sufficiently large bubbles all reactors may reach prompt criticality.
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| 2. |
- Andersson, Peter, 1981-, et al.
(författare)
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Correction for dynamic bias error in transmission measurements of void fraction
- 2012
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Ingår i: Review of Scientific Instruments. - : AIP Publishing. - 0034-6748 .- 1089-7623. ; 83:12, s. 125110-
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Tidskriftsartikel (refereegranskat)abstract
- Dynamic bias errors occur in transmission measurements, such as X-ray, gamma, or neutron radiography or tomography. This is observed when the properties of the object are not stationary in time and its average properties are assessed. The nonlinear measurement response to changes in transmission within the time scale of the measurement implies a bias, which can be difficult to correct for. A typical example is the tomographic or radiographic mapping of void content in dynamic two-phase flow systems. In this work, the dynamic bias error is described and a method to make a first-order correction is derived. A prerequisite for this method is variance estimates of the system dynamics, which can be obtained using high-speed, time-resolved data acquisition. However, in the absence of such acquisition, a priori knowledge might be used to substitute the time resolved data. Using synthetic data, a void fraction measurement case study has been simulated to demonstrate the performance of the suggested method. The transmission length of the radiation in the object under study and the type of fluctuation of the void fraction have been varied. Significant decreases in the dynamic bias error were achieved to the expense of marginal decreases in precision.
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| 3. |
- Andersson, Peter, 1981-, et al.
(författare)
-
Design and initial 1D radiography tests of the FANTOM mobile fast-neutron radiography and tomography system
- 2014
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 756, s. 82-93
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Tidskriftsartikel (refereegranskat)abstract
- The FANTOM system is a tabletop sized fast-neutron radiography and tomography system newly developed at the Applied Nuclear Physics Division of Uppsala University. The main purpose of the system is to provide time-averaged steam-and-water distribution measurement capability inside the metallic structures of two-phase test loops for Light Water Reactor thermal-hydraulic studies using a portable fusion neutron generator. The FANTOM system provides a set of 1D neutron transmission data, which may be inserted into tomographic reconstruction algorithms to achieve a 2D mapping of the steam-and-water distribution. In this paper, the selected design of FANTOM is described and motivated. The detector concept is based on plastic scintillator elements, separated for spatial resolution. Analysis of pulse heights on an event-to-event basis is used for energy discrimination. Although the concept allows for close stacking of a large number of detector elements, this demonstrator is equipped with only three elements in the detector and one additional element for monitoring the yield from the neutron generator. The first measured projections on test objects of known configurations are presented. These were collected using a Sodern Genie 16 neutron generator with an isotropic yield of about 1E8 neutrons per second, and allowed for characterization of the instrument’s capabilities. At an energy threshold of 10 MeV, the detector offered a count rate of about 500 cps per detector element. The performance in terms of spatial resolution was validated by fitting a Gaussian Line Spread Function to the experimental data, a procedure that revealed a spatial unsharpness in good agreement with the predicted FWHM of 0.5 mm.
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| 4. |
- Andersson, Peter, 1981-, et al.
(författare)
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Effects of proton escape on detection efficiency in thin scintillator elements and its consequences for optimization of fast-neutron imaging
- 2011
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 651:1, s. 110-116
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Tidskriftsartikel (refereegranskat)abstract
- Plastic scintillators are commonly used for neutron detection in the MeV energy range, based on n–p scattering and the subsequent deposition of recoil proton's kinetic energy in the detector material. This detection procedure gives a quasi-rectangular energy deposition distribution for mono-energetic neutrons, extending from zero to the neutron energy. However, if the detector sensitive element (DSE) is small, the energy deposition may be incomplete due to the recoil proton escape.In the application of neutron imaging, here exemplified by fast-neutron tomography, two conflicting requirements have been identified: (1) thin DSEs are required to obtain high spatial resolution and (2) energy discrimination may be required to reduce the influence of neutrons being scattered into the DSEs, which generally occurs at lower energies. However, at small DSE widths, the reduction of energy deposition due to recoil proton escape may cause a significant decrease in detection efficiency when energy discrimination is applied.In this work, energy deposition distributions in small-size DSEs have been simulated for Deuterium–Deuterium (DD; 2.5 MeV) and Deuterium–Tritium (DT; 14.1 MeV) fusion neutrons. The intrinsic efficiency has been analyzed as a function of energy discrimination level for various detector widths. The investigations show that proton recoil escape causes a significant drop in intrinsic detection efficiency for thin DSEs. For DT neutrons, the drop is 10% at a width of 3.2 mm and 50% at a width of 0.6 mm, assuming an energy threshold at half the incident neutron energy. The corresponding widths for a DD detector are 0.17 and 0.03 mm, respectively.Finally, implications of the proton escape effect on the design of a fast-neutron tomography device for void distribution measurements at Uppsala University are presented. It is shown that the selection of DSE width strongly affects the instrument design when optimizing for image unsharpness.
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| 5. |
- Andersson, Peter, 1981-
(författare)
-
Fast-Neutron Tomography using a Mobile Neutron Generator for Assessment of Steam-Water Distributions in Two-Phase Flows
- 2014
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- This thesis describes the measurement technique of fast-neutron tomography for assessing spatial distributions of steam and water in two-phase flows. This so-called void distribution is of importance both for safe operation and for efficient use of the fuel in light water reactors, which compose the majority of the world’s commercial nuclear reactors. The technique is aimed for usage at thermal-hydraulic test loops, where heated two-phase flows are being investigated under reactor-relevant conditions.By deploying portable neutron generators in transmission tomography, the technique becomes applicable to stationary objects, such as thermal-hydraulic test loops. Fast neutrons have the advantage of high transmission through metallic structures while simultaneously being relatively sensitive to the water/void content. However, there are also challenges, such as the relatively low yield of commercially available fast-neutron generators, the tendency of fast neutrons to scatter in the interactions with materials and the relatively low efficiency encountered in fast-neutron detection.The thesis describes the design of a prototype instrument, FANTOM, which has been assembled and demonstrated. The main design parameters have been optimized to achieve maximal signal count rate in the detector elements, while simultaneously reaching an image unsharpness of ≤0.5 mm. Radiographic projections recorded with the assembled instrument are presented, and the performance parameters of FANTOM are deduced.Furthermore, tomographic reconstruction methods for axially symmetric objects, which is relevant for some test loops, have been developed and demonstrated on measured data from three test objects. The attenuation distribution was reconstructed with a radial resolution of 0.5 mm and an RMS error of 0.02 cm-1, based on data recorded using an effective measurement time of 3.5 hours per object. For a thermal-hydraulic test loop, this can give a useful indication of the flow mode, but further development is desired to improve the precision of the measurements.Instrument upgrades are foreseen by introducing a more powerful neutron generator and by adding detector elements, speeding up the data collection by several orders of magnitude and allowing for higher precision data. The requirements and performance of an instrument for assessment of arbitrary non-symmetric test loops is discussed, based on simulations.
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| 6. |
- Andersson, Peter, 1981-, et al.
(författare)
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Neutron tomography for void distribution measurements
- 2010
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Ingår i: ENC 2010 Transactions. - 9789295064096 ; , s. 40-45
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Neutron tomography has previously been performed using large, stationary neutron sources such as reactors and spallation sources for applications where the object under study can be transported to the source. This paper accounts for the challenges met when applying neutron tomography using a portable accelerator driven neutron generator, which is required when studying non-transportable objects. In general, portable sources offer significantly lower neutron yields than stationary sources, implying the need for either longer measurement times or highly efficient measurement and/or analysis procedures.The particular application investigated here is the mapping of steam distributions in water (void distribution), which is of high importance for the performance of nuclear fuel assemblies in boiling water reactors (BWR). The void distribution cannot be measured directly in a reactor core, so instead various electrically-heated thermal-hydraulic test loops are used. In these loops, void correlations can be determined in full-size fuel-assembly models, such as FRIGG in Sweden and DESIRE in Holland, but measurements are also performed in smaller, less complicated geometries. Previously, gamma tomography has been used to measure the void distribution in the FRIGG loop. However, improved capabilities to map the void distribution can be expected using neutrons because of their higher sensitivity to water relative to metal structures, as compared to gamma rays. At the same time, neutrons as probe also give rise to some challenges, such as high background from scattering.This paper investigates the possibility to use neutron tomography at axially symmetric objects such as the HWAT test loop in Sweden, where an annular two-phase flow of water/void is confined and heated by a steel cylinder. Monte Carlo simulations of the HWAT geometry and a suggested measurement setup have been carried out, using the particle transport code MCNPX. A reconstruction technique which exploits the symmetries in the test loop has been developed, making it possible to reconstruct the internal void distribution from one single projection. A reconstruction is presented, which is based on simulated data corresponding to a 13-min measurement using a DT source emitting 2∙109 neutrons/s. The reconstruction offers a radial view of the local void fraction in 10 annular sections of HWAT, with uncertainties between 2 and 5 void percent units.
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| 7. |
- Andersson, Peter, 1981-, et al.
(författare)
-
Neutron tomography of axially symmetric objects using 14 MeV neutrons from a portable neutron generator
- 2014
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Ingår i: Review of Scientific Instruments. - : AIP Publishing. - 0034-6748 .- 1089-7623. ; 85:8, s. 085109-
-
Tidskriftsartikel (refereegranskat)abstract
- In nuclear boiling water reactor cores, the distribution of water and steam (void) is essential for both safety and efficiency reasons. In order to enhance predictive capabilities, void distribution assessment is performed in two-phase test-loops under reactor-relevant conditions. This article proposes the novel technique of fast-neutron tomography using a portable deuterium-tritium neutron generator to determine the void distribution in these loops.Fast neutrons have the advantage of high transmission through the metallic structures and pipes typically concealing a thermal-hydraulic test loop, while still being fairly sensitive to the water/void content. However, commercially available fast-neutron generators also have the disadvantage of a relatively low yield and fast-neutron detection also suffers from relatively low detection efficiency. Fortunately, some loops are axially symmetric, a property which can be exploited to reduce the amount of data needed for tomographic measurement, thus limiting the interrogation time needed.In this article, three axially-symmetric test objects depicting a thermal-hydraulic test loop have been examined; steel pipes with outer diameter 24 mm, thickness 1.5 mm and with three different distributions of the plastic material POM inside the pipes. Data recorded with the FANTOM fast-neutron tomography instrument have been used to perform tomographic reconstructions to assess their radial material distribution. Here, a dedicated tomographic algorithm that exploits the symmetry of these objects has been applied, which is described in the paper.Results are demonstrated in 20 rixel (radial pixel) reconstructions of the interior constitution and 2D visualization of the pipe interior is demonstrated. The local POM attenuation coefficients in the rixels were measured with errors (RMS) of 0.025, 0.020 and 0.022 cm-1, solid POM attenuation coefficient. The accuracy and precision is high enough to provide a useful indication on the flow mode, and a visualization of the radial material distribution can be obtained. A benefit of this system is its potential to be mounted at any axial height of a two-phase test section without requirements for pre-fabricated entrances or windows. This could mean a significant increase in flexibility of the void distribution assessment capability at many existing two-phase test loops.
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| 8. |
- Andersson, Peter, 1981-
(författare)
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Optimization of Equipment for Tomographic Measurements of Void Distributions using Fast Neutrons
- 2011
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- This licentiate thesis describes a novel nondestructive measuring technique for determiningspatial distributions of two-phase water flows. In Boiling Water Reactors, which compose themajority of the world's commercial nuclear reactors, this so called void distribution is of importance for safe operation.The presented measurement technique relies on fast neutron transmission tomography using portable neutron generators. Varying hardware options for such an instrument based on this technique and a prototype instrument, which is under construction, are described. The main design parameters are detailed and motivated from a performance point of view. A Paretomultiple objective optimization of the count rate and image unsharpness is presented. The resulting instrument design comprises an array of plastic scintillators for neutron detection. Such detector elements allow for spectroscopic data acquisition and subsequent reduction of background events at low energy by means of introducing an energy threshold in the analysis.The thesis includes two papers: In paper I, the recoil proton energy deposition distribution resulting from the interaction of the incoming neutrons is investigated for thin plastic scintillator elements. It is shown that the recoil proton losses have a large effect on the pulse height distribution and the intrinsic neutron detection efficiency is calculated for varying energy thresholds.In paper II the performance of the planned FANTOM device is investigated using the particle transport code MCNP5. An axially symmetric phantom void distribution is modeled and there construction is compared with the correct solution. According to the solutions, the phantom model can be reconstructed with 10 equal size ring-shaped picture elements, with a precision of better than 5 void percent units using a deuterium-tritium neutron generator with a yield of 3 · 107 neutrons per second and a measurement time of 13 h. However, it should be noted that commercial neutron generators with a factor of 103 higher yields exist and that the measurement time could decrease to less than a minute if such a neutron generator would beutilized.
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| 9. |
- Branger, Erik, 1988-, et al.
(författare)
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Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies
- 2018
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Ingår i: Journal of Instrumentation. - 1748-0221. ; 13
-
Tidskriftsartikel (refereegranskat)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a tool used by nuclear safeguards inspectors to verify irradiated nuclear fuel assemblies in wet storage based on the recording of Cherenkov light produced by the assemblies. One type of verification involves comparing the measured light intensity from an assembly with a predicted intensity, based on assembly declarations. Crucial for such analyses is the performance of the prediction model used, and recently new modelling methods have been introduced to allow for enhanced prediction capabilities by taking the irradiation history into account, and by including the cross-talk radiation from neighbouring assemblies in the predictions.In this work, the performance of three models for Cherenkov-light intensity prediction is evaluated by applying them to a set of short-cooled PWR 17x17 assemblies for which experimental DCVD measurements and operator-declared irradiation data was available; (1) a two-parameter model, based on total burnup and cooling time, previously used by the safeguards inspectors, (2) a newly introduced gamma-spectrum-based model, which incorporates cycle-wise burnup histories, and (3) the latter gamma-spectrum-based model with the addition to account for contributions from neighbouring assemblies.The results show that the two gamma-spectrum-based models provide significantly higher precision for the measured inventory compared to the two-parameter model, lowering the standard deviation between relative measured and predicted intensities from 15.2% to 8.1% respectively 7.8%.The results show some systematic differences between assemblies of different designs (produced by different manufacturers) in spite of their similar PWR 17x17 geometries, and possible ways are discussed to address such differences, which may allow for even higher prediction capabilities. Still, it is concluded that the gamma-spectrum-based models enable confident verification of the fuel assembly inventory at the currently used detection limit for partial defects, being a 30% discrepancy between measured and predicted intensities, while some false detection occurs with the two-parameter model. The results also indicate that the gamma-spectrum-based prediction methods are accurate enough that the 30% discrepancy limit could potentially be lowered.
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| 10. |
- Branger, Erik, 1988-, et al.
(författare)
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Experimental study of background subtraction in Digital Cherenkov Viewing Device measurements
- 2018
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Ingår i: Journal of Instrumentation. - 1748-0221. ; 13:8
-
Tidskriftsartikel (refereegranskat)abstract
- The Digital Cherenkov Viewing Device (DCVD) is an imaging tool used by authority inspectors for partial defect verification of nuclear fuel assemblies in wet storage, i.e. to verify that part of an assembly has not been diverted. One of the currently adopted verification procedures is based on quantitative measurements of the assembly's Cherenkov light emissions, and comparisons to an expected intensity, calculated based on operator declarations. A background subtraction of the intensity data in the recorded images is necessary for accurate quantitative measurements. The currently used background subtraction is aimed at removing an electronics-induced image-wide offset, but it is argued here that the currently adopted procedure may be insufficient.It is recommended that a standard dark-frame subtraction should be used, to remove systematic pixel-wise background due to the electronics, replacing the currently used offset procedure. Experimental analyses show that a dark-frame subtraction would further enhance the accuracy and reliability of DCVD measurements. Furthermore, should ageing of the CCD chip result in larger systematic pixel-wise deviations over time, a dark-frame subtraction can ensure reliable measurements regardless of the age of the CCD chip. It can also help in eliminating any adverse effects of malfunctioning pixels. In addition to the background from electronic noise, ways to compensate for background from neighbouring fuel assemblies and ambient light are also discussed.
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| 11. |
- Branger, Erik, 1988-, et al.
(författare)
-
Image analysis as a tool for improved use of the Digital Cherenkov Viewing Device for inspection of irradiated PWR fuel assemblies.
- 2014
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a tool used to measure the Cherenkov light emitted from irradiated nuclear fuel assemblies stored in water pools. It has been approved by the IAEA for attended gross defect verification, as well as for partial defect verification, where a fraction of the fuel material has been diverted. In this report, we have investigated the current procedures for recording images with the DCVD, and have looked into ways to improve these procedures. Using three different image sets of PWR fuel assemblies, we have analysed what information and results can be obtained using image analysis techniques. We have investigated several error sources that distort the images, and have shown how these errors affect the images. We have also described some of the errors mathematically, and have discussed how these error sources may be compensated for, if the character and magnitude of the errors are known. Resulting from our investigations are a few suggestions on how to improve the procedures and consequently the quality of the images recorded with the DCVD as well as suggestions on how to improve the analysis of collected images. Specifically, a few improvements that should be looked into in the short term are:• Images should be recorded with the fuel assembly perfectly centered in the image, and preferably without any tilt of the DCVD relative to the fuel in order to obtain accurate measurements of the light intensity. Image analysis procedures that may aid the alignment are presented.• To compensate for the distorting effect of the water surface and possible turbulence in the water, several images with short exposure time should be captured rather than one image with long exposure time. Using image analysis procedures, it is possible to sum the images resulting in a final image with less distortions and improved quality.• A reference image should be used to estimate device-related distortions, so that these distortions are compensated for. Ideally, this procedure can also be used to calibrate individual pixels.• The background should be carefully taken into account in order to separate the background level from diffuse signal components, allowing for the background to be subtracted. Accordingly, each measurement campaign should be accompanied by at least one background measurement, recorded from a section in the storage pool where no fuel assemblies are present. Furthermore, the background level should be determined from a larger region in the image and not from one individual pixel, as is currently done.• A database of measurements should be set up, containing DCVD images, information about the applied DCVD settings and the conditions that the DCVD was used in. Any partial defect verification procedure at any time could then be tested against as much data as possible. Accordingly, a database can aid in evaluating and improving partial defect verification methods using DCVD image analysis.Based on the findings and discussions in this report, some long-term improvements are also suggested.
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| 12. |
- Branger, Erik, 1988-, et al.
(författare)
-
Improved DCVD assessments of irradiated nuclear fuel using image analysis techniques
- 2014
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a tool for measuring the Cherenkov light intensity emitted from irradiated nuclear fuel in wet storage. It is currently used in nuclear facilities where authority inspectors perform attended gross defect verification to ensure the presence of irradiated fuel material, as well as partial defect verification to ensure that a fraction of the fuel material has not been diverted. In 2013, Uppsala University (UU), supported by the Swedish Radiation Safety Authority, initiated a PhD project aimed at gaining a better understanding of the underlying physics process of the Cherenkov light emission and its detection, in order to improve and enhance the capabilities of the DCVD. The scope of this research is broad and includes modelling, simulations and experiments. As a first step, expertise on image analysis was brought into the project with the purpose to identify image analysis related opportunities and challenges relevant to the DCVD. The investigations performed so far cover general aspects of image analysis as well as aspects specific for verification of PWR fuels, where the fuel geometry may be extra challenging. Resulting from the investigation are suggestions on how to improve the measurement procedure and consequently the image quality obtained with the DCVD. This presentation describes these results and expected outcomes of their implementation.
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| 13. |
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| 14. |
- Branger, Erik, 1988-, et al.
(författare)
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On the inclusion of light transport in prediction tools for Cherenkov light intensity assessment of irradiated nuclear fuel assemblies
- 2019
-
Ingår i: Journal of Instrumentation. - 1748-0221. ; 14
-
Tidskriftsartikel (refereegranskat)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a tool used to verify irradiated nuclear fuel assemblies in wet storage by imaging the Cherenkov light produced by the radiation emitted from the assemblies. It is frequently used for partial defect verification, verifying that part of an assembly has not been removed and/or replaced. In one of the verification procedures used, the detected total Cherenkov light intensities from a set of assemblies are compared to predicted intensities, which are calculated using operator declarations for the assemblies.This work presents a new, time-efficient method to simulate DCVD images of fuel assemblies, allowing for estimations of the Cherenkov light production, transport and detection. Qualitatively, good agreement between simulated and measured images is demonstrated. Quantitatively, it is shown that relative intensity predictions based on simulated images are within 0.5% of corresponding predictions based solely on the production of Cherenkov light, neglecting light transport and detection. Consequently, in most cases it is sufficient to use predictions based on produced Cherenkov light, neglecting transport and detection, thus substantially reducing the time needed for simulations.In a verification campaign, assemblies are grouped according to their type, and the relative measured and predicted intensities are compared in a group. By determining transparency factors, describing the fraction of Cherenkov light that is blocked by the top plate of an assembly, it is possible to adjust predictions based on the production of Cherenkov light to take the effect of the top plate into account. This procedure allows assemblies of the same type bit with different top plates to be compared with increased accuracy. The effect of using predictions adjusted with transparency factors were assessed experimentally on a set of Pressurized Water Reactor 17x17 assemblies having five different top plate designs. As a result of the adjustment, the agreement between measured and predicted relative intensities for the whole data set was enhanced, resulting in a reduction of an RMSE from 14.1% to 10.7%. It is expected that further enhancements may be achieved by introducing more detailed top-plate and spacer descriptions.
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| 15. |
- Branger, Erik, 1988-, et al.
(författare)
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Towards unattended partial-defect verification of irradiated nuclear fuel assemblies using the DCVD
- 2014
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a tool used by authority inspectors to verify irradiated nuclear fuel assemblies in wet storage by measuring the Cherenkov light emitted. The DCVD is approved by the IAEA for gross defect verification, and is one of the few inspection tools approved for partial defect verification.There is interest in adapting the DCVD to work in unattended mode, so that it can be used to verify large quantities of irradiated fuel assemblies prior to moving them to difficult-to-access storage locations. This work presents methods based on image analysis that can be used to reduce the effects of different types of distortions encountered when performing measurements with the DCVD. Implementing these methods will ensure that data of high quality is obtained. Verification prior to moving fuels to difficult-to-access storage may also require a dedicated measurement station to be built, and it is argued that by constructing these stations with the DCVD in mind, many distortions can be reduced or eliminated. Thus, by implementing safeguards-by-design, it is possible to ensure that the DCVD is used in near optimal conditions.
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| 16. |
- 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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| 17. |
- Davour, Anna, 1975-, et al.
(författare)
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Image analysis methods for partial defect detection using tomographic images on nuclear fuel assemblies
- 2015
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- A promising non-destructive assay method for verification of irradiated nuclear fuel is gammatomography, i.e. the use of measurements of the gamma radiation field around a nuclear fuel assembly to reconstruct detailed information about the internal source distribution.Typically, tomographic reconstructions result in two-dimensional images of cross sections of the fuel. We demonstrate how such images can be searched for fuel rods using a template matching technique, which is a method commonly used in the field of image analysis. In this case, a template or mask corresponding to the size and shape of a fuel rod is translated across the image in order to find the region with the highest reconstructed activity, which is assumed to correspond to the location of a fuel rod in the image. This is done iteratively, allowing no overlap of the rods. By defining the threshold between background and fuel rod objects in the image, we can identify and count the fuel rods using no other assumptions than the rod radius.Thus the rod identification procedure provides a possible means to verify whether all fuel rods arepresent, and it may also be implemented to identify the fuel type of the measured assembly. Theprocedure is robust in cases of irregularities, such as assembly bow or torsion, or the dislocation ofindividual fuel rods in the measured cross section.Here we demonstrate fuel rod identification procedure, using authentic images collected with a tomographic measurement device on commercial fuel assemblies. The results show that image analysis can support tomographic partial defect verification of irradiated nuclear fuel assemblies, even on the single fuel rod level.
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| 18. |
- Grape, Sophie, 1982-, et al.
(författare)
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Forskning inom teknisk kärnämneskontroll vid Uppsala universitet under 2014–2015
- 2016
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- Uppsala universitet har inom ramen för olika avtal med SSM under 2014-2015 bedrivit ett omfattande forskningsprogram inom kärnämneskontroll. Forskningsprogrammet har under denna tid innefattat 3 doktorander med dedikerade forskningsprojekt och ett flertal seniora forskare som helt eller delvis har varit engagerade inom kärnämneskontroll.Denna rapport uppmärksammar särskilt fyra forskningsområden av hög relevans för den globala kärnämneskontrollen, vilka benämns; DCVD, Next Generation Safeguards Initiative, verifiering av atypiska bränsleobjekt och Generation IV kärnkraftsystem. Även andra forskningsaktiviteter har genomförts inom ramen för forskningsprogrammet, vilka dock ligger utanför redovisningen i denna rapport.Under perioden 2014-2015 producerades inom forskningsprogrammet 9 artiklar som skickats till vetenskapliga tidskrifter med peer-review-granskning. Därutöver gjordes medvetna satsningar på att lyfta fram forskningen på de arenor som är av störst betydelse för det internationella kärnämneskontrollarbetet, d.v.s. på de symposier och möten som arrangeras av FN:s internationella atomenergiorgan (IAEA), det europeiska samarbetsorganet ESARDA och den amerikanska organisationen INMM. Vid dessa internationella konferenser publicerades ytterligare 15 vetenskapliga artiklar med unikt innehåll under perioden. En publikationslista med samtliga forskningsarbeten som producerats under perioden redovisas i denna rapport.
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| 19. |
- Grape, Sophie, et al.
(författare)
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Modelling Cherenkov light from irradiated nuclear fuel assemblies using GEANT4
- 2010
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is currently used by International Atomic Energy Agency (IAEA) inspectors for gross defect verification of spent nuclear fuel assemblies in storage pools. A Cherenkov light image is obtained from the spent fuel and the verification is made by the detection of unique Cherenkov characteristics of spent fuel. To take further advantage of its quantitative capabilities, the DCVD’s ability to detect partial defects down to the 30% level is now being investigated. To evaluate the performance of the DCVD, simulations of the emitted and recorded light can be very useful. This presentation describes how the software toolkit GEANT4 is used to gain better understanding of the light contributions from the fuel and its environment by means of Monte Carlo simulations. The toolkit allows the user to access information on individual photon emission coordinates and their momentum vectors and it is also possible to take the expected rod-by-rod burnup distribution at different axial levels into account. Investigations have shown that the Cherenkov light production about the fuel is dominated by gamma radiation from the fuel material interacting with the water surrounding the fuel. A study of the range of the Cherenkov photon production from individual fuel rods, which is of relevance for partial-defect verification, is presented. In addition, emission distributions of Cherenkov light are presented for simulated PWR fuel assemblies with different configurations of replaced rods. Simulated light intensities in guide tubes are presented, showing variations depending on whether fuel rods nearby have been substituted or not.
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| 20. |
- Grape, Sophie, 1982-, et al.
(författare)
-
New perspectives on nuclear power - Generation IV nuclear energy systems to strengthen nuclear non-proliferation and support nuclear disarmament
- 2014
-
Ingår i: Energy Policy. - : Elsevier BV. - 0301-4215 .- 1873-6777. ; 73, s. 815-819
-
Tidskriftsartikel (refereegranskat)abstract
- Recently, nuclear power has received support from environmental and climate researchers emphasizing the need to address factors of global importance such as climate change, peace and welfare. Here, we add to previous discussions on meeting future climate goals while securing safe supplies of energy by discussing future nuclear energy systems in the perspective of strengthening nuclear non-proliferation and aiding in the process of reducing stockpiles of nuclear weapons materials.New nuclear energy systems, currently under development within the Generation IV (Gen IV) framework, are being designed to offer passive safety and inherent means to mitigate consequences of nuclear accidents. Here, we describe how these systems may also be used to reduce or even eliminate stockpiles of civil and military plutonium—the former present in waste from today׳s reactors and the latter produced for weapons purposes. It is argued that large-scale implementation of Gen IV systems would impose needs for strong nuclear safeguards. The deployment of Safeguards-by-Design principles in the design and construction phases can avoid draining of IAEA resources by enabling more effective and cost-efficient nuclear safeguards, as compared to the current safeguards implementation, which was enforced decades after the first nuclear power plants started operation.
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| 21. |
- Grape, Sophie, et al.
(författare)
-
Partial Defect Evaluation Methodology for Nuclear Safeguards Inspections of Used Nuclear Fuel Using the Digital Cherenkov Viewing Device
- 2014
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 186:1, s. 90-98
-
Tidskriftsartikel (refereegranskat)abstract
- This paper describes possible ways of analyzing and interpreting data obtained using the digital Cherenkov viewing device on spent nuclear fuel assemblies for the identification of partial defects in the fuel. According to the terminology of the International Atomic Energy Agency, partial defects refer to items, for instance, fuel assemblies, that are manipulated to the extent that a fraction of the fuel material is diverted or substituted. Analysis can be performed either by using a measure of the total light intensity or by identifying the light distribution pattern emanating from the spent nuclear fuel, the goal of either type of analysis being a quantitative measure that can be used in the data interpretation step. Two possible data interpretation alternatives are presented here: the threshold method and the hypothesis testing method. This paper summarizes some of the simulation studies and results that have been obtained, related to the two analysis and data interpretation methodologies.
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| 22. |
- Grape, Sophie, 1982-, et al.
(författare)
-
Partial defect verification using the DCVD : a capability evaluation approach
- 2011
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is a non-intrusive instrument available to theInternational Atomic Energy Agency (IAEA) for verifying spent nuclear fuel in storage pools. It iscurrently used for gross-defect evaluations, i.e. to verify that an item in a storage pool is anirradiated fuel assembly and not a fresh assembly or a dummy. This is done by recording images ofthe Cherenkov light emitted in the water surrounding the fuel. Currently, the instrument’s ability toalso detect partial defects at the 50% level or even lower is under study. Here, experimental work iscomplimented by modeling and simulations due to the limited availability of assemblies with partialdefects.Ideally, an IAEA inspector should be able to use the DCVD at e.g. a fuel storage site andimmediately after scanning obtain information on (1) whether an item is an irradiated fuel assemblyor not, and (2) whether the assembly is intact or suffers from a partial defect. This paper discusses adecision-making methodology intended for the latter purpose with the objective to implement it inthe DCVD software in order to facilitate smooth inspection procedures. Inspectors will thus not berequired to possess any expertise in the decision-making methodology.The paper also describes measurements performed during spring 2011 at the CLAB interim spentfuel storage in Sweden. The measurements were carried out with the objective to optimize theequipment handling and work flow during this type of measurement campaigns and to form a basisfor the evaluation of the DCVD’s ability to detect partial defects.
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| 23. |
- Grape, Sophie, 1982-, et al.
(författare)
-
Recent modelling studies for analysing the partial-defect detection capability of the Digital Cherenkov Vieweing Device
- 2014
-
Ingår i: Esarda Bulletin. - Ispra. - 0392-3029. ; :51, s. 3-8
-
Tidskriftsartikel (refereegranskat)abstract
- Strong sources of radioactivity, such as spent nuclear fuel stored in water pools, give rise to Cherenkov light. This light originates from particles, in this case electrons released from gamma-ray interactions, which travel faster than the speed of light in the water. In nuclear safeguards, detection of the Cherenkov light intensity is used as a means for verifying gross and partial defect of irradiated fuel assemblies in wet storage. For spent nuclear fuel, the magnitude of the Cherenkov light emission depends on the initial fuel enrichment (IE), the power history (in particular the total fuel burnup (BU)) and the cooling time (CT). This paper presents recent results on the expected Cherenkov light emission intensity obtained from modelling a full 8x8 BWR fuel assembly with varying values of IE, BU and CT. These results are part of a larger effort to also investigate the Cherenkov light emission for fuels with varying irradiation history and other fuel geometries in order to increase the capability to predict the light intensity and thus lower the detection limits for the Digital Cherenkov Viewing Device (DCVD). The results show that there is a strong dependence of the Cherenkov light intensity on BU and CT, in accordance with previous studies. However, the dependences demonstrated previously are not fully repeated; the current study indicates a less steep decrease of the intensity with increasing CT. Accordingly, it is suggested to perform dedicated experimental studies on fuel with different BU and CT to resolve the differences and to enhance future predictive capability. In addition to this, the dependence of the Cherenkov light intensity on the IE has been investigated. Furthermore, the modelling of the Cherenkov light emission has been extended to CTs shorter than one year. The results indicate that high-accuracy predictions for short-cooled fuel may require more detailed information on the irradiation history.
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| 24. |
- Grape, Sophie, 1982-, et al.
(författare)
-
Recent modelling studies for analysing the partial-defect detection capability of theDigital Cherenkov Viewing Device
- 2013
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is an instrument available to IAEA inspectors forverifying spent nuclear fuel in wet storage at nuclear facilities. The instrument records the Cherenkovlight that is emitted in the water surrounding the highly radioactive fuel. The light intensity is largelydependent on the amount of nuclear material in the fuel as well as its burnup and cooling time and can beused by the inspector as a measure for verifying the properties of the fuel.To aid in the analysis of the Cherenkov light intensity, a simulation toolkit has been developed, whichmodels the emission, transport and detection of Cherenkov light. This toolkit is particularly useful forinvestigating the response of the DCVD for fuel assemblies subject to different types of partial defects,where fuel rods might have been removed or substituted with non-irradiated material. Variousconfigurations of partial defects may be simulated in order to evaluate the detection capabilities of theDCVD.Here, we present how the light intensity recorded by the DCVD is affected by the fuel history and by thepartial defect scenario. We present a methodology for how the analysis and interpretation of recordedintensities may be performed to result in confidence-supported statements of different levels of partialdefect. Finally, we suggest topics for further studies to accomplish an automated inspection system based on this methodology.
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| 25. |
- Grape, Sophie, 1982-, et al.
(författare)
-
Students’ approaches to learning from other students’ oral presentations
- 2013
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- A phenomenographic study has been performed in order to investigate students’ approaches to learning from other students’ oral presentations in the context of a compulsory seminar on nuclear accidents in the third year of the nuclear engineering programme at Uppsala University.
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| 26. |
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| 27. |
- Grape, Sophie, 1982-, et al.
(författare)
-
Verifying nuclear fuel assemblies in wet storages on a partial defect level : A software simulation tool for evaluating the capabilities of the Digital Cherenkov Viewing Device
- 2013
-
Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 698, s. 66-71
-
Tidskriftsartikel (refereegranskat)abstract
- The Digital Cherenkov Viewing Device (DCVD) is an instrument that records the Cherenkov light emitted from irradiated nuclear fuels in wet storages. The presence, intensity and pattern of the Cherenkov light can be used by the International Atomic Energy Agency (IAEA) inspectors to verify that the fuel properties comply with declarations. The DCVD is since several years approved by the IAEA for gross defect verification, i.e. to control whether an item in a storage pool is a nuclear fuel assembly or a non-fuel item [1]. Recently, it has also been endorsed as a tool for partial defect verification, i.e. to identify if a fraction of the fuel rods in an assembly have been removed or replaced. The latter recognition was based on investigations of experimental studies on authentic fuel assemblies and of simulation studies on hypothetic cases of partial defects [2]. This paper describes the simulation methodology and software which was used in the partial defect capability evaluations. The developed simulation procedure uses three stand-alone software packages: the ORIGEN-ARP code [3] used to obtain the gamma-ray spectrum from the fission products in the fuel, the Monte Carlo toolkit Geant4 [4] for simulating the gamma-ray transport in and around the fuel and the emission of Cherenkov light, and the ray-tracing programme Zemax [5] used to model the light transport through the assembly geometry to the DCVD and to mimic the behaviour of its lens system. Furthermore, the software allows for detailed information from the plant operator on power and/or burnup distributions to be taken into account to enhance the authenticity of the simulated images. To demonstrate the results of the combined software packages, simulated and measured DCVD images are presented. A short discussion on the usefulness of the simulation tool is also included
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| 28. |
- Hellesen, Carl, et al.
(författare)
-
Improved proliferation resistance of fast reactor blankets manufactured from spent nuclear fuel
- 2013
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- In this paper we investigate how a blanket manufactured from recycled light water reactor (LWR)waste, instead of depleted uranium (DU), could potentially improve the non- proliferationcharacteristics. The blanket made from LWR waste would from the start of operation contain a fractionof plutonium isotopes unsuitable for weapons production. As 239Pu is bred in the blanket it istherefore always mixed with the plutonium already present.We use a Monte Carlo model of the advanced burner test reactor (ABTR) as reference design, andthe proliferation resistance of the blanket material is evaluated for two criteria, spontaneous neutronemission and decay heat. We show that it is possible to achieve a production of plutonium withproliferation resistance comparable to light water reactor waste with a burnup of 50MWd/kg.
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| 29. |
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| 30. |
- Hellesen, Carl, et al.
(författare)
-
Transient Simulation of Gas Bubble in a Medium Sized Lead Cooled Fast Reactor
- 2014
-
Ingår i: Proceedings of the International Conference on Physics of Reactors (PHYSOR 2014).
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- A common problem for many liquid metal cooled fast reactor designs is the positive void worth of the coolant. In this context, an advantage of lead cooled fast reactors is the high temperature of coolant boiling. In contrast to sodium cooled fast reactors this, in practice, precludes coolant boiling. However, partial voiding of the core could result from e.g. gas bubbles entering the core from below. This would introduce a positive reactivity, if the bubble is large enough. In this paper we model this type of event using a point kinetics code coupled to a heat transport code. The reactivity parameters are obtained from a Monte Carlo code. The 300 MWth reactor design Alfred is used as a test case. We show that in general the reactor design studied is robust in such events, and we conclude that small bubbles a measureable Power oscillation would occur. For very large bubbles there exist a possibility of core damage. The cladding is the most sensitive part.
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| 31. |
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| 32. |
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| 33. |
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| 34. |
- Holcombe, Scott, et al.
(författare)
-
A Novel gamma emission tomography instrument for enhanced fuel characterization capabilities within the OECD Halden Reactor Project
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 837-845
-
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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| 35. |
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| 36. |
- Holcombe, S., et al.
(författare)
-
Advanced fuel assembly characterization capabilities based on gamma tomography at the halden boiling water reactor
- 2012
-
Ingår i: Proc. Int. Conf. on Advances in Reactor Physics. - 9781622763894 ; , s. 3478-3489
-
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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| 37. |
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| 38. |
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| 39. |
- Holcombe, Scott, et al.
(författare)
-
Feasibility of identifying leaking fuel rods using gamma tomography
- 2013
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 57, s. 334-340
-
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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| 40. |
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| 41. |
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| 42. |
- Holcombe, Scott
(författare)
-
Gamma Spectroscopy and Gamma Emission Tomography for Fuel Performance Characterization of Irradiated Nuclear Fuel Assemblies
- 2014
-
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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| 43. |
- Holcombe, Scott, et al.
(författare)
-
Method For Analyzing Fission Gas Release In Fuel Rods Based On Gamma-Ray Measurements Of Short-Lived Fission Products
- 2013
-
Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 184:1, s. 96-106
-
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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| 44. |
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| 45. |
- Holcombe, Scott
(författare)
-
Nondestructive Fission Gas Measurements by Means of Gamma Spectroscopy and Gamma Tomography
- 2012
-
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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| 46. |
- Jacobsson Svärd, Staffan, 1970-
(författare)
-
A Tomographic Measurement Technique for Irradiated Nuclear Fuel Assemblies
- 2004
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The fuel assemblies used at the Swedish nuclear power plants contain typically between 100 and 300 fuel rods. An experimental technique has been demanded for determining the relative activities of specific isotopes in individual fuel rods without dismantling the assemblies. The purpose is to validate production codes, which requires an experimental relative accuracy of <2 % (1 σ).Therefore, a new, non-destructive tomographic measurement technique for irradiated nuclear fuel assemblies has been developed. The technique includes two main steps: (1) the gamma-ray flux distribution around the assembly is recorded, and (2) the interior gamma-ray source distribution in the assembly is reconstructed. The use of detailed gamma-ray transport calculations in the reconstruction procedure enables accurate determination of the relative rod-by-rod source distribution.To investigate the accuracy achievable, laboratory equipment has been constructed, including a fuel model with a well-known distribution of 137Cs. Furthermore, an instrument has been constructed and built for in-pool measurements on irradiated fuel assemblies at nuclear power plants.Using the laboratory equipment, a relative accuracy of 1.2 % was obtained (1 σ). The measurements on irradiated fuel resulted in a repeatability of 0.8 %, showing the accuracy that can be achieved using this instrument. The agreement between rod-by-rod data obtained in calculations using the POLCA–7 production code and measured data was 3.1 % (1 σ).Additionally, there is a safeguards interest in the tomographic technique for verifying that no fissile material has been diverted from fuel assemblies, i.e. that no fuel rods have been removed or replaced. The applicability has been demonstrated in a measurement on a spent fuel assembly. Furthermore, detection of both the removal of a rod as well as the replacement with a non-active rod has been investigated in detail and quantitatively established using the laboratory equipment.
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| 47. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
-
Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies
- 2015
-
Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier. - 0168-9002 .- 1872-9576. ; 783, s. 128-141
-
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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| 48. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
-
Detection of Dislocated Individual Fuel Rods in a Nuclear Fuel Assembly using Tomographic Measurements
- 1998
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A method is suggested for identifying and quantifying possible dislocations of individual fuel rods in an irradiated nuclear fuel assembly. The method is designed for application in tomographic measurements of nuclear fuel assemblies. The source distribution of gamma radiation is reconstructed using a tomographic algorithm, in which the pixel pattern is adapted to the assembly geometry. By comparing the reconstructed source concentration in opposite parts of each fuel rod in the assembly, quantitative information may be obtained about possible dislocations.Theoretical considerations have been applied and data from simulations of a nuclear fuel assembly with single dislocated rods have been used in tomographic reconstructions. The investigations indicate that the method should be applicable for identification of dislocations larger than a few tenths of a mm.
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| 49. |
- Jacobsson Svärd, Staffan, 1972-, et al.
(författare)
-
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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- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Modeling of the Cherenkov Light Emission from Nuclear Fuel Assemblies with Partial Defects
- 2010
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Ingår i: PHYSOR 2010. - La Grange Park, Illinois, USA : American Nuclear Society. - 9780894480799
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Konferensbidrag (refereegranskat)abstract
- The International Atomic Energy Agency poses requirements on the detection and verification of partial defects of nuclear fuel assemblies before being placed in difficult-to-access storage. One instrument being considered for such detection is the Digital Cherenkov Viewing Device, with which images of the Cherenkov light from fuel assemblies in storage pools can be recorded and analyzed. This paper accounts for a software toolkit for simulating the Cherenkov photon distribution in the fuel using GEANT4. The toolkit enables the user to access information on individual photon emission coordinates and their momentum vectors, as well as to take into account the expected rod-by-rod burnup distribution at different axial levels. An example of this modeling is demonstrated.
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