| 1. |
- Branger, Erik, 1988-, et al.
(författare)
-
Comparison of prediction models for Cherenkov light emissions from nuclear fuel assemblies
- 2017
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Ingår i: Journal of Instrumentation. - 1748-0221. ; 12
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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 Cherenkov light produced by the assembly. Verification that no rods have been substituted in the fuel, so-called partial-defect verification, is made by comparing the intensity measured with a DCVD with a predicted intensity, based on operator fuel declaration. The prediction model currently used by inspectors is based on simulations of Cherenkov light production in a BWR 8x8 geometry. This work investigates prediction models based on simulated Cherenkov light production in a BWR 8x8 and a PWR 17x17 assembly, as well as a simplified model based on a single rod in water. Cherenkov light caused by both fission product gamma and beta decays were considered.The simulations reveal that there are systematic differences between the models, most noticeably with respect to the fuel assembly cooling time. Consequently, a prediction model that is based on another fuel assembly configuration than the fuel type being measured, will result in systematic over or underestimation of short-cooled fuel as opposed to long-cooled fuel. While a simplified model may be accurate enough for fuel assemblies with fairly homogeneous cooling times, the prediction models may differ by up to 18 \,\% for more heterogeneous fuel. Accordingly, these investigations indicate that the currently used model may need to be exchanged with a set of more detailed, fuel-type specific models, in order minimize the model dependant systematic deviations.
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| 2. |
- Branger, Erik, 1988-, et al.
(författare)
-
Experimental evaluation of models for predicting Cherenkov light intensities from short-cooled nuclear fuel assemblies
- 2018
-
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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| 3. |
- Branger, Erik, 1988-, et al.
(författare)
-
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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| 4. |
- Branger, Erik, 1988-, et al.
(författare)
-
Improved Cherenkov Light Prediction Model for Enhanced DCVD Performance
- 2018
-
Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is an instrument used to verify irradiated nuclear fuel assemblies in wet storage based on the fuel’s Cherenkov light emissions. The DCVD is frequently used for partial defect verification, verifying that 50% or more of an assembly has not been diverted. The verification methodology is based on comparison of the measured Cherenkov light intensity to a predicted intensity, based on operator declarations.For the last five years, a dedicated PhD project at Uppsala University has been aiming at enhancing and improving the verification capabilities when using the DCVD. The project is now approaching its end, and this paper summarizes the comprehensive work performed regarding improving the prediction capabilities.A new prediction model has been developed, considering more fuel assembly details to ensure more accurate predictions. With the new model, the irradiation history of an assembly, the assembly design and the contributions from gamma and beta decays are taken into account. The model has also been extended to account for the radiation from neighbouring fuel assemblies, which can enter the assembly being measured and contribute to the measured Cherenkov light. The performance of the prediction model and the neighbour intensity prediction model has been validated against fuel measurements by the IAEA at a PWR facility with short-cooled fuel. The results show that the new model offers an improved prediction capability, allowing the fuel inventory to be verified with no fuel assemblies being identified as outliers requiring additional investigation. A simplified version of the prediction model will be implemented in the next DCVD software version, making it available to IAEA inspectors.This development of the DCVD capabilities are in line with the fourth theme of the IAEA safeguards symposium, “Shaping the future of safeguards implementation”, by resolving challenges related to the DCVD and by extending the capabilities of the instrument.
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| 5. |
- Branger, Erik, 1988-, et al.
(författare)
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Investigating the Cherenkov light production due to cross-talk in closely stored nuclear fuel assemblies in wet storage
- 2018
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Ingår i: ESARDA Bulletin. - : European Commission Joint Research Centre. - 1977-5296. ; :57, s. 66-74
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Tidskriftsartikel (övrigt vetenskapligt/konstnärligt)abstract
- The Digital Cherenkov Viewing Device (DCVD) is one of the tools available to a safeguards inspector performing verifications of irradiated nuclear fuel assemblies in wet storage. One of the main advantages of safeguards verification using Cherenkov light is that it can be performed without moving the fuel assemblies to an isolated measurement position, allowing for quick measurements. One disadvantage of this procedure is that irradiated nuclear fuel assemblies are often stored close to each other, and consequently gamma radiation from one assembly can enter a neighbouring assembly, and produce Cherenkov light in the neighbour. As a result, the measured Cherenkov light intensity of one assembly will include contributions from its neighbours, which may affect the safeguards conclusions drawn.In this paper, this so-called near-neighbour effect, is investigated and quantified through simulation. The simulations show that for two fuel assemblies with similar properties stored closely, the near-neighbour effect can cause a Cherenkov light intensity increase of up to 3% in a measurement. For one fuel assembly surrounded by identical neighbour assemblies, a total of up to 14% of the measured intensity may emanate from the neighbours. The relative contribution from the near-neighbour effect also depends on the fuel properties; for a long-cooled, low-burnup assembly, with low gamma and Cherenkov light emission, surrounded by short-cooled, high-burnup assemblies with high emission, the measured Cherenkov light intensity may be dominated by the contributions from its neighbours.When the DCVD is used for partial-defect verification, a 50% defect must be confidently detected. Previous studies have shown that a 50% defect will reduce the measured Cherenkov light intensity by 30% or more, and thus a threshold has been defined, where a ≥30% decrease in Cherenkov light indicates a partial defect. However, this work shows that the near-neighbour effect may also influence the measured intensity, calling either for a lowering of this threshold or for the intensity contributions from neighbouring assemblies to be corrected for. In this work, a method is proposed for assessing the near-neighbour effect based on declared fuel parameters, enabling the latter type of corrections.
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| 6. |
- Branger, Erik, 1988-, et al.
(författare)
-
On Cherenkov light production by irradiated nuclear fuel rods
- 2017
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Ingår i: Journal of Instrumentation. - 1748-0221. ; 12
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Tidskriftsartikel (refereegranskat)abstract
- Safeguards verification of irradiated nuclear fuel assemblies in wet storage is frequently done by measuring the Cherenkov light in the surrounding water produced due to radioactive decays of fission products in the fuel. This paper accounts for the physical processes behind the Cherenkov light production caused by a single fuel rod in wet storage, and simulations are presented that investigate to what extent various properties of the rod affect the Cherenkov light production. The results show that the fuel properties has a noticeable effect on the Cherenkov light production, and thus that the prediction models for Cherenkov light production which are used in the safeguards verifications could potentially be improved by considering these properties.It is concluded that the dominating source of the Cherenkov light is gamma-ray interactions with electrons in the surrounding water. Electrons created from beta decay may also exit the fuel and produce Cherenkov light, and e.g. Y-90 was identified as a possible contributor to significant levels of the measurable Cherenkov light in long-cooled fuel. The results also show that the cylindrical, elongated fuel rod geometry results in a non-isotropic Cherenkov light production, and the light component parallel to the rod's axis exhibits a dependence on gamma-ray energy that differs from the total intensity, which is of importance since the typical safeguards measurement situation observes the vertical light component. It is also concluded that the radial distributions of the radiation sources in a fuel rod will affect the Cherenkov light production.
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| 7. |
- Branger, Erik, 1988-, et al.
(författare)
-
On the inclusion of light transport in prediction tools for Cherenkov light intensity assessment of irradiated nuclear fuel assemblies
- 2019
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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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| 8. |
- 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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| 9. |
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| 10. |
- Hellesen, Carl, 1980-, et al.
(författare)
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Nuclear Spent Fuel Parameter Determination using Multivariate Analysis of Fission Product Gamma Spectra
- 2017
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 886-895
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Tidskriftsartikel (refereegranskat)abstract
- In this paper, we investigate the application of multivariate data analysis methods to the analysis of gamma spectroscopy measurements of spent nuclear fuel (SNF). Using a simulated irradiation and cooling of nuclear fuel over a wide range of cooling times (CT), total burnup at discharge (BU) and initial enrichments (IE) we investigate the possibilities of using a multivariate data analysis of the gamma ray emission signatures from the fuel to determine these fuel parameters. This is accomplished by training a multivariate analysis method on simulated data and then applying the method to simulated, but perturbed, data.We find that for SNF with CT less than about 20 years, a single gamma spectrum from a high resolution gamma spectrometer, such as a high-purity germanium spectrometer, allows for the determination of the above mentioned fuel parameters.Further, using measured gamma spectra from real SNF from Swedish pressurized light water reactors we were able to confirm the operator declared fuel parameters. In this case, a multivariate analysis trained on simulated data and applied to real data was used.
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| 11. |
- Jacobsson, Staffan, 1970-
(författare)
-
Changes in the relative pin power distribution in a nuclear fuel assembly due to channel bow and pin dislocations
- 2003
-
Rapport (övrigt vetenskapligt/konstnärligt)abstract
- Simulations of the relative pin power distribution in a nuclear fuel assembly have been performed using the core-analysis code CASMO-4. A cross section of a previously unirradiated BWR assembly of the GE12S fuel type was simulated in a burnup range of 0–16 MWd/kgU. The simulated void content was kept constant at 25%. Two main types of geometric disturbances from the nominal assembly geometry were investigated: (1) channel bow and (2) dislocations of individual fuel pins. The disturbances were simulated to be constant throughout the whole burnup range.It was concluded that the first type of disturbance could give rise to the largest changes in relative pin power, as compared to the non-disturbed case. The maximum increase was about 4% per simulated mm channel bow up to a simulated bow of 9 mm. Due to the reflective boundary conditions used in CASMO‑4, this corresponds to a 2% change in pin power per mm change in water gap between adjacent assemblies. For dislocations of individual fuel pins, the largest increase in relative pin power was 2.6% per mm, obtained for a peripheral pin. The largest changes were generally obtained at beginning of cycle (BOC).As expected due to effects from enhanced neutron moderation, it was found that relative pin powers generally increased in regions where water gaps were widened and vice versa. There was also an influence from BA-pins, i.e. pins with a content of burnable neutron absorbers. When a pin was dislocated towards a BA-pin, its relative power decreased. The decrease in BA content with irradiation also gave rise to non-linear dependencies between burnup and changes in BA pin power, as compared to the non-disturbed case.
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| 12. |
- Jacobsson, Staffan, 1970-, et al.
(författare)
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Outcomes of the JNT 1955 Phase I Viability Study of Gamma Emission Tomography for Spent Fuel Verification
- 2017
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Ingår i: ESARDA Bulletin. - 1977-5296. ; :55, s. 10-28
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Tidskriftsartikel (refereegranskat)abstract
- The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly has been assessed within the IAEA Support Program project JNT 1955, phase I, which was completed and reported to the IAEA in October 2016. Two safeguards verification objectives were identified in the project; (1) independent determination of the number of active pins that are present in a measured assembly, in the absence of a priori information about the assembly; and (2) quantitative assessment of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives was evaluated across a range of fuel types, burnups and cooling times, while targeting a total interrogation time of less than 60 minutes.The evaluations were founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of "virtual" fuel assemblies. The simulated instrument response data were then processed using a variety of tomographic-reconstruction and image- processing methods, and scoring metrics were defined and used to evaluate the performance of the methods.This paper describes the analysis framework and metrics used to predict tomographer performance. It also presents the design of a "universal" GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA. Finally, it gives examples of the expected partial-defect detection capabilities for some fuels and diversion scenarios, and it provides a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.
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| 13. |
- Jacobsson, Staffan, 1970-, et al.
(författare)
-
Tomographic measurements of thermal power in nuclear fuel rods : Stage 2 progress report December 1999
- 1999
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report presents recent progress in a project on tomographic measurements of thermal power in nuclear fuel rods, carried out at Uppsala University and funded via the Swedish Centre for Nuclear Technology, KTC. The project is executed in three stages, of which this report describes a set of studies made during the second stage.Experimental studies have been performed using a laboratory mock-up, modelling a fuel assembly of the BWR8x8 type, in which tomographic data collection is made using BGO scintillator detectors and a data-acquisition system based on single-channel analysers. Gamma-ray scattering has been identified as a major contributor to systematic errors in the measurement of relative activity contents in the 63 rods of the mock-up assembly. Since scattering causes build-up of radiation at lower energies, it may be taken into account in the tomographic analyses by introducing a so-called effective attenuation coefficient in the reconstruction models, being slightly lower than the theoretical coefficient. Studies show that this approach may enhance the precision in the measurement of relative rod-activity contents from about 3‑4% down to about 1.2‑1.4%.Data collection has also been performed using a separate, spectroscopic data-acquisition system, in a set of measurements where inactive rods have been used to introduce scattering in order to analyse its effects on the collected data. The results indicate that most of the adverse effects of scattering may be eliminated by deploying a spectroscopic system with peak analysis including background subtraction. Consequently, such a system should be considered for Stage 3 of this project.Simulation studies have also been executed to analyse the measurement uncertainties introduced by geometric deviations from the nominal positions of the four sections of a SVEA‑64 fuel assembly. It was found that the standard deviation of relative rod activities caused by the largest displacements allowed by the nominal gaps enclosing each section was about 0.5%, which may be considered acceptable.
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| 14. |
- Jacobsson Svärd, Staffan, 1970-
(författare)
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A Tomographic Measurement Technique for Irradiated Nuclear Fuel Assemblies
- 2004
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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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| 15. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies
- 2015
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier. - 0168-9002 .- 1872-9576. ; 783, s. 128-141
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Tidskriftsartikel (refereegranskat)abstract
- A fuel assembly operated in a nuclear power plant typically contains 100–300 fuel rods, depending on fuel type, which become strongly radioactive during irradiation in the reactor core. For operational and security reasons, it is of interest to experimentally deduce rod-wise information from the fuel, preferably by means of non-destructive measurements. The tomographic SPECT technique offers such possibilities through its two-step application; (1) recording the gamma-ray flux distribution around the fuel assembly, and (2) reconstructing the assembly׳s internal source distribution, based on the recorded radiation field. In this paper, algorithms for performing the latter step and extracting quantitative relative rod-by-rod data are accounted for.As compared to application of SPECT in nuclear medicine, nuclear fuel assemblies present a much more heterogeneous distribution of internal attenuation to gamma radiation than the human body, typically with rods containing pellets of heavy uranium dioxide surrounded by cladding of a zirconium alloy placed in water or air. This inhomogeneity severely complicates the tomographic quantification of the rod-wise relative source content, and the deduction of conclusive data requires detailed modelling of the attenuation to be introduced in the reconstructions. However, as shown in this paper, simplified models may still produce valuable information about the fuel.Here, a set of reconstruction algorithms for SPECT on nuclear fuel assemblies are described and discussed in terms of their quantitative performance for two applications; verification of fuel assemblies׳ completeness in nuclear safeguards, and rod-wise fuel characterization. It is argued that a request not to base the former assessment on any a priori information brings constraints to which reconstruction methods that may be used in that case, whereas the use of a priori information on geometry and material content enables highly accurate quantitative assessment, which may be particularly useful in the latter application.Two main classes of algorithms are covered; (1) analytic filtered back-projection algorithms, and (2) a group of model-based or algebraic algorithms. For the former class, a basic algorithm has been implemented, which does not take attenuation in the materials of the fuel assemblies into account and which assumes an idealized imaging geometry. In addition, a novel methodology has been presented for introducing a first-order correction to the obtained images for these deficits; in particular, the effects of attenuation are taken into account by modelling the response for an object with a homogeneous mix of fuel materials in the image area. Neither the basic algorithm, nor the correction method requires prior knowledge of the fuel geometry, but they result in images of the assembly׳s internal activity distribution. Image analysis is then applied to deduce quantitative information.Two algebraic algorithms are also presented, which model attenuation in the fuel assemblies to different degrees; either assuming a homogenous mix of materials in the image area without a priori information or utilizing known information of the assembly geometry and of its position in the measuring setup for modelling the gamma-ray attenuation in detail. Both algorithms model the detection system in detail. The former algorithm returns an image of the cross-section of the object, from which quantitative information is extracted, whereas the latter returns conclusive relative rod-by-rod data.Here, all reconstruction methods are demonstrated on simulated data of a 96-rod fuel assembly in a tomographic measurement setup. The assembly was simulated with the same activity content in all rods for evaluation purposes. Based on the results, it is argued that the choice of algorithm to a large degree depends on application, and also that a combination of reconstruction methods may be useful. A discussion on alternative analysis methods is also included.
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| 16. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Detection of Dislocated Individual Fuel Rods in a Nuclear Fuel Assembly using Tomographic Measurements
- 1998
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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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| 17. |
- 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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| 18. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Tomografisk bestämning av termisk effekt i kärnbränslestavar - Förstudie
- 1996
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- Denna rapport redovisar en förstudie av möjligheterna att med tomografisk mätning bestämma effektfördelningen i ett bränsleelement. Förstudien avses vara en första fas i en serie av tre, som, om utfallet successivt bedöms tillräckligt lovande, skall leda till en utprovad prototyp för tomografisk effektmätning.
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| 19. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Tomographic determination of spent fuel assembly pin-wise burnup and cooling time for detection of anomalies
- 2015
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The IAEA has initiated Member States’ Support Program project JNT A 1955 to assess the partial defect detection capabilities of gamma emission tomography (GET) for spent nuclear fuel assembly verification. The GET technique is based on measurements of the gamma-ray flux distribution around a spent fuel assembly using dedicated, tomographic equipment and subsequent reconstruction of the internal source distribution using tomographic algorithms applied on the recorded data. One of the verification objectives identified for the project is the quantitative measurement of pin-by-pin properties, e.g. burnup and/or cooling time, for the detection of anomalies and/or verification of operator-declared data. For this objective, reconstruction algorithms that return quantitative, isotopic pin-by-pin data are applied.Previously, GET measurements performed on commercial nuclear fuel assemblies in Sweden have proven capable of determining the relative pin-by-pin power distribution with high precision in BWR fuel with short cooling time, based on the measured distribution of the gamma-ray emitting fission product 140Ba/La in the fuel. In the current project, the capabilities of GET to determine additional pin-wise fuel parameters in additional fuel types are being assessed. The evaluations are based on Monte Carlo simulations of the emission of gamma-rays from the fuel and their detection in a tomographic measurement device.This paper describes the algorithms used for reconstructing quantitative pin-wise data and the results that are anticipated with this technique. It is argued that detailed modelling of the gamma-ray attenuation through the highly inhomogeneous mix of strongly-attenuating fuel rods and less-attenuating surrounding water (wet storage) or air (dry storage) is required to yield high precision in the reconstructed data. The burnup distribution assessment would be based on the recording of 662-keV gamma radiation from 137Cs, whereas the assessment of both burnup and cooling time simultaneously requires the GET measurement and pin-wise reconstruction of at least two isotopes, which puts constraints on the measurement equipment used.
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| 20. |
- Jacobsson Svärd, Staffan, 1970-, et al.
(författare)
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Verification of completeness of spent nuclear fuel assemblies by means of tomography
- 2008
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Ingår i: International Conference on the Physics of Reactors. - Switzerland : Paul Scherrer Institute. - 9783952140956
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Within the safeguards regime, new technologies are desired to allow for detection of possible diversion of individual fuel rods from an irradiated nuclear fuel assembly. A measurement technique, based on Single-Photon Emission Computed Tomography, has been developed and demonstrated on a SVEA‑96 BWR assembly, which had been irradiated for one power cycle at the Forsmark 2 NPP. Images of the assembly’s internal distribution of the gamma-ray emitting isotope 140Ba/140La have been obtained tomographically, without requiring the assembly to be dismantled. Image analysis has been applied and the capability to detect the removal of a single rod from the assembly centre has been demonstrated. In addition, rod-activity reconstructions have been performed, involving detailed modelling of the gamma-ray transport through the fuel. An agreement of 3.1% (1 s) with data from the production code POLCA‑7 was demonstrated. An inspection procedure is suggested where tomographic data is collected, online image reconstruction is performed and image analysis is applied, resulting in a preliminary statement of the assembly’s completeness less than a minute after the measurement is completed. In cases where the completeness is questionable, an off-line rod-activity reconstruction can be performed, resulting in highly accurate data without the need for additional measurements.
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| 21. |
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| 22. |
- Jansson, Peter, 1971-, et al.
(författare)
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A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly
- 2006
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 152:1, s. 76-86
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Tidskriftsartikel (refereegranskat)abstract
- There is a general interest in experimentally determining the power distribution in nuclear fuel. The prevalent method is to measure the distribution of the fission product 140Ba, which represents the power distribution over the last few weeks. In order to obtain the rod-by-rod power distribution, the fuel assemblies have to be dismantled.In this paper, a device for experimental nondestructive determination of the thermal rod-by-rod power distribution in boiling water reactor and pressurized water reactor fuel assemblies is described. It is based on measurements of the 1.6-MeV gamma radiation from the decay of 140Ba/La and utilizes a tomographic method to reconstruct the rod-by-rod source distribution. No dismantling of the fuel assembly is required.The device is designed to measure an axial node in 20 min with a precision of >2% (1). It is primarily planned to be used for validation of production codes for core simulation but may also be used for safeguards purposes.
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| 23. |
- Jansson, Peter, 1971-, et al.
(författare)
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Tomographic measurement of thermal power in nuclear fuel : Stage 1 - Final report
- 1998
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- This report summarises the results obtained in the first stage of a three-stage project on “Tomographic measurement of thermal power in nuclear fuel”, carried out at Uppsala University and funded via the Swedish Centre for Nuclear Technology, KTC. The goal of the three-stage project is to build a tomographic device and use it for non-destructive measurement of the pin-wise thermal power in nuclear fuel assemblies at a Swedish nuclear power plant.Covered in the report are descriptions of the algorithms and software for tomographic reconstruction developed for the purpose, results of sensitivity studies for geometric uncertainties, tests of detector systems in a laboratory environment as well as on nuclear fuel at a reactor site, descriptions of a laboratory mock-up intended to be used in the second step of the project and estimations of the achievable uncertainty in measured relative pin-wise power using the suggested methods.
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| 24. |
- Lundqvist Saleh, Tobias, et al.
(författare)
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Recent Progress in the Design of a Tomographic Device for Measurements of the Three-Dimensional Pin-Power Distribution in Irradiated Nuclear Fuel Assemblies
- 2010
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Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 165:2, s. 232-239
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Tidskriftsartikel (refereegranskat)abstract
- A tomographic technique for determination of the thermal power distribution in nuclear fuel assemblies is under development. The purpose is to provide an experimental validation tool for core simulation codes. Such codes are essential for the operation of nuclear power reactors, and validation is important in the process of improving and developing the codes as well as the fuel. The tomographic method is nonintrusive and offers large amounts of data within a normal revision shutdown. In earlier experimental investigations using a test platform, the method proved useful, demonstrating results of satisfying quality. However, the measuring setup also revealed nonfeasible properties related to transport, decontamination, and background radiation shielding. In this paper, the design of a new measuring device is presented. It is based on experiences from the test platform, but its size and weight make it advantageous regarding transports and decontamination. Moreover, the design inherently allows for more efficient background shielding. The latter has been investigated in a detailed study using the MCNP simulation code. The results confirm the high levels of background radiation observed in the test platform. It is also concluded that the shielding properties in the new design are sufficient.
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| 25. |
- Smith, Eric L., et al.
(författare)
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A Viability Study of Gamma Emission Tomography for Spent Fuel Verification : JNT 1955 Phase I Technical Report
- 2016
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Rapport (refereegranskat)abstract
- The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclear fuel assembly is being assessed through a collaboration of Support Programs to the International Atomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectives have been identified. The first is the independent determination of the number of active pins that are present in the assembly, in the absence of a priori information about the assembly. The second objective is to provide quantitative assay of pin-by-pin properties, for example the activity of key isotopes or pin attributes such as cooling time and relative burnup, under the assumption that basic fuel parameters (e.g., assembly type and nominal fuel composition) are known. The efficacy of GET to meet these two verification objectives has been evaluated across a range of fuel types, burnups, and cooling times, and with a target total interrogation time of less than 60 minutes. This evaluation of GET viability for safeguards applications was founded on a modelling and analysis framework applied to existing and emerging GET instrument designs. Monte Carlo models of different fuel types were used to produce simulated tomographer responses to large populations of “virtual” fuel assemblies. Instrument response data were processed using a variety of tomographic-reconstruction and image-processing methods, and scoring metrics specific to each of the verification objectives were used to predict performance. This report describes the analysis framework and metrics used to predict tomographer performance, the design of a “universal” GET (UGET) instrument intended to support the full range of verification scenarios envisioned by the IAEA, and a comparison of predicted performance for the notional UGET design and an optimized variant of an existing IAEA instrument.
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| 26. |
- Verma, Vasudha, 1988-, et al.
(författare)
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Self Powered Neutron Detectors as in-core detectors for Sodium-cooled Fast Reactors
- 2017
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 860, s. 6-12
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Tidskriftsartikel (refereegranskat)abstract
- Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction.In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.
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