| 1. |
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| 2. |
- 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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| 3. |
- 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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| 4. |
- Grape, Sophie, 1982-, et al.
(författare)
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New perspectives on nuclear power - Generation IV nuclear energy systems to strengthen nuclear non-proliferation and support nuclear disarmament
- 2014
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Ingår i: Energy Policy. - : Elsevier BV. - 0301-4215 .- 1873-6777. ; 73, s. 815-819
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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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| 5. |
- Grape, Sophie, 1982-, et al.
(författare)
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Students’ approaches to learning from other students’ oral presentations
- 2013
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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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| 6. |
- Hellesen, Carl, et al.
(författare)
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Improved proliferation resistance of fast reactor blankets manufactured from spent nuclear fuel
- 2013
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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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| 7. |
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| 8. |
- Jacobsson Svärd, Staffan, 1972-, et al.
(författare)
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Gamma-ray Emission Tomography: Modelling and evaluation of partial-defect testing capabilities
- 2014
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Assessment of gamma emission tomography (GET) for spent nuclear fuel verification is the task in IAEA MSP project JNT1955. In line with IAEA Safeguards R&D plan 2012-2023, the aim of this effort is to “develop more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect tests on spent fuel assemblies prior to transfer to difficult to access storage". The current viability study constitutes the first phase of three, with evaluation and decision points between each phase. Two verification objectives have been identified; (1) counting of fuel pins in tomographic images without any a priori knowledge of the fuel assembly under study, and (2) quantitative measurements of pin-by-pin properties, e.g. burnup, for the detection of anomalies and/or verification of operator-declared data.Previous measurements performed in Sweden and Finland have proven GET highly promising for detecting removed or substituted fuel pins (i.e. partial defects) in BWR and VVER-440 fuel assemblies even down to the individual fuel pin level. The current project adds to previous experiences by pursuing a quantitative assessment of the capabilities of GET for partial defect detection, across a broad range of potential IAEA applications, fuel types, and fuel parameters. A modelling and performance-evaluation framework has been developed to provide quantitative GET performance predictions, incorporating burn-up and cooling-time calculations, Monte Carlo radiation-transport and detector-response modelling, GET instrument definitions (existing and notional) and tomographic reconstruction algorithms, which use recorded gamma-ray intensities to produce cross-sectional images of the source distribution in the fuel assembly or conclusive pin-by-pin data. The framework also comprises image-processing algorithms and performance metrics that recognize the inherent trade-off between the probability of detecting missing pins and the false-alarm rate. Here, the modelling and analysis framework is described and preliminary results are presented.
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| 9. |
- Jansson, Peter, 1971-, et al.
(författare)
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A laboratory device for developing analysis tools and methods for gamma emission tomography of nuclear fuel
- 2013
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Tomography is a measurement technique that images the inner parts of objects using only external measurement. It is widely used within the field of medicine, and may become important also for nuclear fuel verification where inspectors can obtain information from fuel assemblies’ inner sections without dismantling them.At Uppsala University, Sweden, a laboratory device has been built for investigating the tomographic measurement techniques on nuclear fuel. The device is composed of machinery to position model fuelrods, activated with Cs-137, in a fuel assembly pattern according to the user's choice. The gamma radiation from the model fuel assembly is collimated to a set of detectors that record the radiation intensity in various positions around the fuel model. Reconstruction of the gamma activity distribution within the fuel model is performed off-line.The objective for constructing the laboratory device was to support the development of tomographic techniques for nuclear fuel diagnostics as well as for nuclear safeguards purposes. The device allows for evaluating the performance of different data-acquisition setups, measurement schemes and reconstruction algorithms, since the activity content of each fuel rod is well known.For safeguards purposes, the device is unique in its capability to model various fuel geometries and configurations of partial defects. The latter includes removed, empty and substituted fuel rods. It is well suited for developing tomographic techniques that are optimized for partial defect detection. It also allows for development of analysis tools necessary to quantify detection limits.Here, we describe the capabilities of the laboratory device and elaborate on how the device may be used to support the nuclear safeguards community with the development of unattended gamma emission tomography.
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| 10. |
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| 11. |
- Jansson, Peter, 1971-, et al.
(författare)
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Gamma Transport Calculations for Gamma Emission Tomography on Nuclear Fuel within the UGET Project
- 2014
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The unattended gamma emission tomography (UGET) for spent nuclear fuel verification is an on-going project in the IAEA member states’ support program. In line with the long term R&D plan of the IAEA Department of Safeguards, it is anticipated that this effort will help develop “more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect test on spent fuel assembly prior to transfer to difficult to access storage”.In the first phase of the project, gamma transport calculations and modelling of exist- ing and proposed new designs of tomographic instruments is performed. In this paper, a set of Monte Carlo calculations regarding modelling of various tomographic devices are presented, including two existing tomographic instruments previously used for spent fuel measurements; one instrument based on scintillator detectors, developed by Uppsala University, and another based on CdTe detector arrays, developed by the JNT 1510 col- laborative effort (Hungary, Finland). Detailed models of the tomographic instruments, including structural materials, and the measured fuel assemblies are used in the simula- tions. The calculated results are compared to the experimentally measured data to provide a benchmark for the simulation procedure.The developed modelling capabilities are also used for evaluation of the partial-defect detection capabilities of the tomographic technique based on a proposed GET instrument design.
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| 12. |
- Martinik, Tomas, et al.
(författare)
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Characterization of Spent Nuclear Fuel with a Differential Die-Away Instrument
- 2014
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Differential die-away instrument (DDA) is currently being investigated within the Next Generation Safeguards Initiative Spent Fuel project as one of the non-destructive assay techniques for spent nuclear fuel characterization and verification. In this paper we report on the progress of designing the first prototype to be deployed at Swedish central interim storage facility (CLAB) where a first set of measurements of 25 PWR and 25BWR spent fuel assemblies is proposed. We also present several working concepts of how the instrument can be customized for dedicated purposes, be it a light weight instrument for portable applications, a minimalist design for reliable and economic operations or a so-called “defectoscope” design for detailed inspection of spent nuclear fuel assemblies.
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| 13. |
- Martinik, Tomas, et al.
(författare)
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Simulation of differential die-away instrument's response to asymmetrically burned spent nuclear fuel
- 2015
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Ingår i: Nuclear Instruments and Methods in Physics Research Section A. - : Elsevier BV. - 0168-9002 .- 1872-9576. ; 788, s. 79-85
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Tidskriftsartikel (refereegranskat)abstract
- Previous simulation studies of differential die-away (DDA) instrument's response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.The results of this study suggest that DDA instrument response depends on the position of the individual neutron detectors and in fact can be split in two modes.The first mode, measured by the back detectors, is not significantly sensitive to the spatial distribution of fissile isotopes and neutron absorbers, but rather reflects the total amount of both contributors as in the cases of symmetrically burned SFAs. In contrary, the second mode, measured by the front detectors, yields certain sensitivity to the orientation of the asymmetrically burned SFA inside the assaying instrument. This study thus provides evidence that the DDA instrument can potentially be utilized as necessary in both ways, i.e. a quick determination of the average SFA characteristics in a single assay, as well as a more detailed characterization involving several DDA observables through assay of the SFA from all of its four sides that can possibly map the burn-up distribution and/or identify diversion or replacement of pins.
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| 14. |
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| 15. |
- Tobin, Stephen, et al.
(författare)
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Experimental and Analytical Plans for the Non-destructive Assay System of the Swedish Encapsulation and Repository Facilities
- 2014
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The Swedish Nuclear Fuel and Waste Management Company (SKB), European Atomic Energy Community (Euratom), two universities and several U.S. Department of Energy Laboratories have joined in a collaborative research effort to determine the capability of non-destructive assay (NDA) techniques to meet the combined needs of the safeguards community and the Swedish encapsulation and repository facilities operator SKB. These needs include partial defect detection, heat quantification, assembly identification (initial enrichment, burnup and cooling time), and Pu mass and reactivity determination. The experimental component of this research effort involves the measurement of 50 assemblies at the Central Storage of Spent Nuclear Fuel (Clab) facility in Sweden, 25 of which were irradiated in Pressurized Water Reactors and 25 in Boiling Water Reactors. The experimental signatures being measured for all assemblies include spectral resolved gammas (HPGe and LaBr3), time correlated neutrons (Differential Die-away Self Interrogation), time-varying and continuous active neutron interrogation (Differential Die-away and an approximation of Californium Interrogation Prompt Neutron), total neutron and total gamma fluxes (Fork Detector), total heat (assembly length calorimeter) and possibly the Cerenkov light emission (Digital Cerenkov Viewing Device). This paper fits into the IAEA’s Department of Safeguards Long-Term R&D Plan in the context of developing “more sensitive and less intrusive alternatives to existing NDA instruments to perform partial defect test on spent fuel assembly prior to transfer to difficult to access storage,” as well as potentially supporting pyrochemical processing. The work describes the specific measured signatures, the uniqueness of the information contained in these signatures and why a data mining approach is being used to combine the various signatures to optimally satisfy the various needs of the collaboration. This paper will address efficient and effective verification strategies particularly in the context of encapsulation and repository facilities.
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| 16. |
- White, Timothy A., et al.
(författare)
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Passive Tomography for Spent Fuel Verification: Analysis Framework and Instrument Design Study
- 2015
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- The potential for gamma emission tomography (GET) to detect partial defects within a spent nuclearfuel assembly is being assessed through a collaboration of Support Programs to the InternationalAtomic Energy Agency (IAEA). In the first phase of this study, two safeguards verification objectiveshave been identified. The first is the independent determination of the number of active pins that arepresent in the assembly, in the absence of a priori information. The second objective is to providequantitative measures of pin-by-pin properties, e.g. activity of key isotopes or pin attributes such ascooling time and relative burnup, for the detection of anomalies and/or verification of operator-declareddata. The efficacy of GET to meet these two verification objectives will be evaluated across a range offuel types, burnups, and cooling times, and with a target interrogation time of less than 60 minutes.The evaluation of GET viability for safeguards applications is founded on a modelling and analysisframework applied to existing and emerging GET instrument designs. Monte Carlo models of differentfuel types are used to produce simulated tomographer responses to large populations of “virtual” fuelassemblies. Instrument response data are processed by a variety of tomographic-reconstruction andimage-processing methods, and scoring metrics specific to each of the verification objectives aredefined and used to evaluate the performance of the methods. This paper will provide a description ofthe analysis framework and evaluation metrics, example performance-prediction results, and describethe design of a “universal” GET instrument intended to support the full range of verification scenariosenvisioned by the IAEA.
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| 17. |
- Wolniewicz, Peter, et al.
(författare)
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Detecting neutron spectrum perturbations due to coolant density changes in a small lead-cooled fast nuclear reactor
- 2013
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 58, s. 102-109
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Tidskriftsartikel (refereegranskat)abstract
- The lead-cooled fast reactor (LFR) is one of the nuclear reactor technologies proposed by the Generation IV International Forum (GIF). The lead coolant allows for inherent safety properties attractive from a nuclear safety point of view, but issues related to corrosion of structural materials and the possible positive coolant reactivity coefficient must be addressed before LFRs can be commercially viable. As an example, a small crack in e.g. a heat exchanger can generate a more or less homogeneous distribution of bubbles in the coolant (void) which if unnoticed, has the potential to cause criticality issues. This fact motivated an investigation of a methodology to detect such voids.The suggested methodology is based on measurements of the “slow” and “fast” parts of the neutron spectrum because these parts respond in different ways to voiding. For detection, it is tentatively assumed that fission chambers loaded with U-235 and Pu-239, respectively, are deployed. To investigate the methodology according to sensitivity and precision, a number of scenarios have been simulated and analysed using the core simulator Serpent.The results show that the methodology yields a sensitivity of 3% for each per cent unit of void. Assuming typical detection limits of a few per cent this implies the possibility to detect voids down to the order of 1%. From these studies it was also concluded that the positioning of the detectors relative the reactor core is crucial, which may be useful input during the design phase of a reactor in order to achieve an efficient monitoring system.
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| 18. |
- Wolniewicz, Peter, et al.
(författare)
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Feasibility study of detection of coolant void in liquid metal cooled fast reactors using changes in the neutron spectrum
- 2013
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 265, s. 1255-1265
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Tidskriftsartikel (refereegranskat)abstract
- Formation of coolant void can lead to an increase in reactivity in metal-cooled fast reactors. Accordingly, the ability to detect formation of void and similar phenomena is highly relevant in order to counteract transient behaviour of such a reactor. As this work shows, the energy distribution of the neutron flux in a fast reactor is sensitive to formation of void. For monitoring purposes, this fact suggests the use of fission chambers with different isotopic content and thus different fission threshold energies. In such a way the monitoring system may be tailored in order to fit the purpose to obtain spectral information of the neutron flux.In this work, simulations have been performed using the Monte-Carlo-based code SERPENT on the ELECTRA reactor design, a 0.5 MWth lead-cooled fast reactor (LFR) planned for in Sweden. The simulations show significant changes in the neutron spectrum due to the formation of void located in specific in-core regions as well as due to a homogeneous core-wide distribution of small bubbles. In an attempt to quantify and to put a number on the spectroscopic changes, the number of neutrons in the high energy region (2–5 MeV) are compared to the number of neutrons in the low-energy region (50–500 keV) and the changes caused by the introduction of void are analyzed. The implications of the findings are discussed.
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