| 1. |
- Andersson, Peter, 1981-
(författare)
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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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| 2. |
- 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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| 3. |
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| 4. |
- Andersson, Peter, 1981-, et al.
(författare)
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Simulation of the response of a segmented High-Purity Germanium detector for gamma emission tomography of nuclear fuel
- 2020
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Ingår i: SN Applied Sciences. - : Springer. - 2523-3963 .- 2523-3971. ; 2
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Tidskriftsartikel (refereegranskat)abstract
- Irradiation testing of nuclear fuel is routinely performed in nuclear test reactors. For qualification and licensing of Accident Tolerant Fuels or Generation IV reactor fuels, an extensive increase in irradiation testing is foreseen in order to fill the gaps of existing validation data, both in normal operational conditions and in order to identify operational limits.Gamma Emission Tomography (GET) has been demonstrated as a viable technique for studies of the behavior of irradiated nuclear fuel, e.g. measurement of fission gas release and inspection of fuel behavior under Loss-Of-Coolant Accident conditions. In this work, the aim is to improve the technique of GET for irradiated nuclear fuel by developing a detector concept for an improved tomography system that allows for a higher spatial resolution and/or faster interrogation.We present the working principles of a novel concept for gamma emission tomography using a segmented High Purity Germanium (HPGe) detector. The performance of this concept was investigated using the Monte Carlo particle transport code MCNP. In particular, the data analysis of the proposed detector was evaluated, and the performance, in terms of full energy efficiency and localization failure rate, has been evaluated.We concluded that the segmented HPGe detector has an advantageous performance as compared to the traditional single-channel detector systems. Due to the scattering nature of gamma rays, a trade-off is presented between efficiency and cross-talk; however, the performance is nevertheless a substantial improvement over the currently used single-channel HPGe detector systems.
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| 5. |
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| 6. |
- Dahlfors, Marcus, 1972-
(författare)
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Studies of Accelerator-Driven Systems for Transmutation of Nuclear Waste
- 2006
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.
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| 7. |
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| 8. |
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| 9. |
- 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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| 10. |
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| 11. |
- Hellesen, Carl, et al.
(författare)
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Transient Simulation of Gas Bubble in a Medium Sized Lead Cooled Fast Reactor
- 2014
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Ingår i: Proceedings of the International Conference on Physics of Reactors (PHYSOR 2014).
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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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| 12. |
- Henriksson, Hans, et al.
(författare)
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Long-term operation of the Swedish Centre for Nuclear Technology (SKC): New challenges and solutions in competence building
- 2016
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Ingår i: Proc. European Conf. Nuclear Education and Training (NESTet 2016), Berlin, Germany, May 22-26, 2016.
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Konferensbidrag (refereegranskat)abstract
- The Swedish Centre for Nuclear Technology (Svenskt Kärntekniskt Centrum, SKC) is a national initiative to perform industry relevant research at Swedish universities, and to support dedicated education of direct use to the Swedish nuclear industry.SKC has been the meeting point between industry and academia for 25 years, and has coped with varying needs from industry and political situations. The present situation in the Nordic countries is split: Sweden plans to shut down four out of ten reactors by 2020, while Finland is planning and constructing new reactors. Even without a strong signal to construct new reactors in Sweden, the need for nuclear competence will stay, as we have challenges in front of us to operate and dismantle power plants, operate the intermediate storage facility CLAB in Oskarshamn, and to build and fill the final repositories in Forsmark.The education supported by the SKC at the selected universities will facilitate possible recruitment for nuclear installations. The funding body of SKC consists of all the Swedish Nuclear Power Plants (NPP) situated in Forsmark (three BWRs), Oskarshamn (three BWRs) and Ringhals (one BWR and three PWRs), and the nuclear fuel manufacturer (Westinghouse), while the main research and education is carried out at Chalmers, KTH Royal Institute of Technology and Uppsala University with corresponding in-kind contribution. The research activities cover highly requested studies for today’s nuclear fleet: material embrittlement, stress-corrosion cracking, accident-tolerant fuel development and ageing management for long-term operation (LTO), while the educational part consists of Bachelor and Master programmes as well as elective courses for students outside the main nuclear programmes and contract education.In the master programmes, focus is on e-learning platforms for courses and examination. Examples from such development are to be presented in the full contribution to the conference. Another success story is project based courses in industry, especially within the Bachelor programmes. This is highly appreciated by students, providing a direct contact with future employers, The success of SKC originates from close contact between the funding bodies and academia on many levels: base funding for course preparation, project support in annual calls, and dedicated long-term research funding. But also dedicated industrial experts, and an enthusiasm from academia to enlighten present research issues, as well as strong presence at universities during student fairs and career days. That is LTO of SKC!
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| 13. |
- Håkansson, Ane, 1959-, et al.
(författare)
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An experimental study of the neutron emission from spent PWR fuel
- 1997
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- Measurements of the thermal and epithermal neutron emission from eleven 15x15 and fourteen 17x17 PWR fuel assemblies have been performed. In the measurements a FORK detector supplied by Euroatom was utilised. The neutron flux was observed to depend on the burnup to approximately the fourth power. Also the strong dependence on initial enrichment could be verified. The latter dependency suggests a possible method to determine the initial enrichment. Such a method is considered as an important feature of safeguard as well as in fuel processing at the planned encapsulation plant for spent nuclear fuel.
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| 14. |
- Håkansson, Ane, 1959-
(författare)
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ANItA : A new Swedish national competence centre in new nuclear power technology
- 2024
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 418
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Tidskriftsartikel (refereegranskat)abstract
- In 2021, a proposal was submitted to the Swedish Energy Agency, which meant that the three leading Swedish universities: Uppsala University, KTH and Chalmers as well as the Swedish and Finnish nuclear power industry would together form a national competence centre for research and development work regarding the introduction of small modular reactors in Sweden. The proposal, called, ANItA (Academic-industrial Nuclear technology Initiative to Achieve a sustainable energy future) entailed a commitment for ten years and was granted funds for the first five years with an option for continued funding for the following five years. The initial research portfolio includes fourteen projects spanning a wide range of technical and non-technical aspects of small modular reactor (SMR) technology. During the last few months, the projects have been staffed with both national and international doctoral students and postdocs and thus constitute a contribution to securing the important supply of expertise in the nuclear field in Sweden. ANItA has attracted great interest from the surrounding society, which includes both politics and mass media and discussions are being held with potential new industrial partners. The renewed interest in new nuclear power in Sweden has further strengthened ANItA's role and the development is moving in a direction that enables to fulfil the basic idea behind ANItA, namely that through research and development work produce knowledge-based decision support for politicians and policymakers in the introduction of new nuclear power technology in Sweden.
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| 15. |
- Håkansson, Ane, et al.
(författare)
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Digital behandling av linjärpulser från en CdTe-detektor : en förstudie
- 1997
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The neccesary treatment of the linear pulses from a CdTe detector in order to improve the energy resolution for gamma-ray spectroscopy is normally performed by using analogue technique. In this paper we suggest two methods based on digital treatment of the detector pulses. Significant features of the methods are the improvement of the energy resolution, the fact that virtually no dead time is introduced in the detector system and the simpler handling of such systems. The paper describes the underlying idea of the methods, computer simulations of detector system and actual measurements. Preliminary results show that an improvement of the energy resolution of a factor of 2 to 5, depending on the method, used is achieved.
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| 16. |
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| 17. |
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| 18. |
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| 19. |
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| 20. |
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| 21. |
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| 22. |
- Jacobsson, Staffan, 1972-, et al.
(författare)
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A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel
- 1998
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A tomographic method for experimental investigation of the integrity of usedLWR fuel has been developed. It is based on measurements of the gamma radiation fromthe fission products in the fuel rods. A reconstruction code of the algebraic type has beenwritten. The potential of the technique has been examined in extensive simulationsassuming a gamma-ray energy of either 0.66 MeV (137Cs) or 1.27 MeV (154Eu).The resultsof the simulations for BWR fuel indicate that single fuel rods or groups of rods replacedwith water or fresh fuel can be reliably detected independent of their position in the fuelassembly using 137Cs radiation. For PWR fuel the same result is obtained with the exceptionof the most central positions. Here the more penetrable radiation from 154Eu must be used inorder to allow a water channel to be distinguished from a fuel rod.The results of the simulations have been verified experimentally for a 8x8 BWRfuel assembly. Special equipment has been constructed and installed at the interim storageCLAB. The equipment allows the mapping of the radiation field around a fuel assemblywith the aid of a germanium detector fitted with a collimator with a vertical slit. Theintensities measured in 2 520 detector positions were used as input for the reconstructioncode used in the simulations. The results agreed very well with the simulations and revealedsignificantly a position containing a water channel in the central part of the assembly.
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| 23. |
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| 24. |
- Jacobsson, Staffan, et al.
(författare)
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A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel Assemblies - I: Simulation Studies
- 2001
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Ingår i: Nuclear Technology. - 0029-5450. ; 135:2, s. 131-145
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Tidskriftsartikel (refereegranskat)abstract
- A tomographic method for experimental investigation of the integrity of used light water reactor fuel assemblies has been developed. It is based on spectroscopic measurements of the gamma radiation from fission products in fuel rods. The method utilizes beforehand information about the nominal geometry of both the measured fuel assembly and the measurement equipment. A reconstruction code of the algebraic type has been written. The potential of the technique has been examined in extensive simulations, assuming a gamma-ray energy of either 662 keV (137Cs) or 1274 keV (154Eu). The ability of detecting various configurations of manipulated rods, both single and in groups, has been investigated. Two main types of manipulations have been simulated. First, there is the removal of rods without replacement. The results indicate that all investigated configurations of removed rods in boiling water reactor (BWR) fuel can be reliably detected using 137Cs radiation. For pressurized water reactor (PWR) fuel, the same result is obtained, with the exception of the most central positions. Here, the more penetrating radiation from 154Eu may have to be used. Second, there is the replacement of rods with fresh fuel or fuel-like material. The results clearly indicate that all simulated cases of such manipulation can be most confidently detected. The simulations include various configurations of replaced rods in both BWR and PWR fuel, using both gamma-ray energies.
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| 25. |
- Jacobsson, Staffan, et al.
(författare)
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A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel Assemblies - II: Experimental Investigation
- 2001
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 135:2, s. 146-153
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Tidskriftsartikel (refereegranskat)abstract
- A tomographic method for verification of the integrity of used light water reactor fuel has been experimentally investigated. The method utilizes emitted gamma rays from fission products in the fuel rods. The radiation field is recorded in a large number of positions relative to the assembly, whereby the source distribution is reconstructed using a special-purpose reconstruction code.An 8 × 8 boiling water reactor fuel assembly has been measured at the Swedish interim storage (CLAB), using installed gamma-scanning equipment modified for the purpose of tomography. The equipment allows the mapping of the radiation field around a fuel assembly with the aid of a germanium detector fitted with a collimator with a vertical slit. Two gamma-ray energies were recorded: 662 keV (137Cs) and 1274 keV (154Eu). The intensities measured in 2520 detector positions were used as input for the tomographic reconstruction code. The results agreed very well with simulations and significantly revealed a position containing a water channel in the central part of the assembly.
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| 26. |
- Jacobsson, Staffan, 1970-, et al.
(författare)
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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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| 27. |
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| 28. |
- 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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| 29. |
- 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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| 30. |
- Jacobsson Svärd, Staffan, et al.
(författare)
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Non-destructive experimental determination of the pin-power distribution in nuclear fuel
- 2003
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- A need for validation of modern core-analysis codes with respect to the calculated pin-power distribution has been recognized. A non-destructive experimental method for such validation has been developed, based on a tomographic technique. Each axial node of the fuel assembly is measured separately and the relative pin-by-pin content of the direct fission product Ba-140 is determined. Investigations performed so far indicate that 1-2% (1 σ) accuracy can be obtained.A measuring device has been constructed which, when fully equipped, is designed to measure a complete BWR assembly in 25 axial nodes within an eight-hour work shift. The applicability of the constructed device has been demonstrated in measurements at the Swedish BWR Forsmark 2 on irradiated fuel with a cooling time of 4-5 weeks. Data from the core-analysis code POLCA-7 have been compared to measured pin-by-pin contents of Ba-140. An agreement of 3.1% (1 σ) has been demonstrated.As compared to the conventional method, involving gamma scanning of individual fuel pins, this method does not require the fuel to be disassembled. Neither does the fuel channel have to be removed. The cost per measured fuel pin is in the order of 20 times lower than the conventional method.
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| 31. |
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| 32. |
- Jacobsson Svärd, Staffan, et al.
(författare)
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Non-Destructive Experimental Determination of the Power Distribution in Nuclear Fuel Assemblies
- 2005
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Ingår i: 2005 International Congress on Advances in Nuclear Power Plants (ICAPP 05).
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Modern production codes for core simulation perform calculations of the thermal power distribution on the individual fuel rod level. The prevalent technique to experimentally validate such calculations involves dismantling of the fuel assemblies, whereby gamma-ray measurements on individual fuel rods are performed to determine the content of the fission product 140Ba. Here, an alternative, non-destructive technique is presented, which is based on tomography. The gamma-ray flux distribution in an axial node is recorded, whereby the relative rod-by-rod content of 140Ba is reconstructed. The method does not require the fuel to be dismantled. Neither does the fuel channel present in BWR assemblies have to be removed. The applicability of the technique has been demonstrated in measurements at the Swedish BWR Forsmark 2 using a special-purpose device. The measurements were performed on irradiated fuel with a cooling time of 4 5 weeks. Data from the production code POLCA 7 has been compared to the measured rod-by-rod contents of 140Ba and an agreement of 3.0% (1 σ) has been demonstrated.
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| 33. |
- Jacobsson Svärd, Staffan, et al.
(författare)
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Nondestructive Experimental Determination of the Pin-Power Distribution in Nuclear Fuel Assemblies
- 2005
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Ingår i: Nuclear Technology. - 0029-5450. ; 151:1, s. 70-76
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Tidskriftsartikel (refereegranskat)abstract
- A need for validation of modern production codes with respect to the calculated pin-power distribution has been recognized. A nondestructive experimental method for such validation has been developed based on a tomographic technique. The gamma-ray flux distribution is recorded in each axial node of the fuel assembly separately, whereby the relative rod-by-rod content of the fission product 140Ba is determined. Measurements indicate that 1 to 2% accuracy (1 sigma) is achievable.A device has been constructed for in-pool measurements at reactor sites. The applicability has been demonstrated in measurements at the Swedish boiling water reactor (BWR) Forsmark 2 on irradiated fuel with a cooling time of 4 to 5 weeks. Data from the production code POLCA-7 have been compared to measured rod-by-rod contents of 140Ba. An agreement of 3.1% (1 sigma) has been demonstrated.It is estimated that measurements can be performed on a complete BWR assembly in 25 axial nodes within an 8-h work shift. As compared to the conventional method, involving gamma scanning of individual fuel rods, this method does not require the fuel to be disassembled nor does the fuel channel have to be removed. The cost per measured fuel rod is estimated to be an order of magnitude lower than the conventional method.
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| 34. |
- 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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| 35. |
- Jacobsson Svärd, Staffan, et al.
(författare)
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Tomographic measurements for partial defect verification – experience with different devices of the stationary type
- 2003
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Ingår i: 25th Annual Meeting - Symposium on Safeguards and Nuclear Materials Management. - 928945654X
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- Tomographic measurements have been performed for the purpose of partial-defect verification on the single-rod level. The measurement procedure involves recording of the gamma radiation field emanating from emitted radiation from within an irradiated assembly and consecutive reconstruction of the internal source distribution. Different devices of the stationary type have been utilised, ranging from a laboratory device used in measurements on a fuel model to an in-pool device used in measurements on irradiated fuel in a fuel-handling pool.The tomographic technique has proven to be robust and reliable. Its applicability for partial-defect verification on the single-rod level has been satisfying. Some required properties of a stationary device are discussed.
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| 36. |
- Jacobsson Svärd, Staffan, et al.
(författare)
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Tomography for partial-defect verification : experiences from measurements using different devices
- 2006
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Ingår i: ESARDA Bulletin. - 0392-3029. ; 33, s. 15-25
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Tidskriftsartikel (refereegranskat)abstract
- Three devices of different types have been used in tomographic measurements for the purpose of partial-defect verification on the single-rod level. The devices range from a laboratory device used in measurements on a fuel model to an in-pool device used in measurements on irradiated fuel in a fuel-handling pool.The tomographic technique accounted for in this paper involves measurements of the gamma-ray flux distribution around a fuel assembly followed by computer-aided reconstruction of the internal source distribution. The results are rod-by-rod values of the relative concentrations of selected gamma-emitting isotopes. Also cross-sectional images are obtained.The tomographic technique presented here has proven to be robust and reliable. In laboratory experiments on a fuel model, reconstructions of relative rod-by-rod activities have been obtained with 1.5 % accuracy (1 σ). Using an in-pool device in measurements on fuel with a cooling time of about 4 weeks, data on fuel rods have been obtained in agreement with production-code calculations. Furthermore, tomographic images of good quality have been acquired.The applicability of the tomographic technique for partial-defect verification on the single-rod level has been investigated and demonstrated. The gamma-ray source concentration reconstructed in a position corresponding to a removed or replaced rod has been significantly lower than that of normal rods.Finally, requirements and properties of a device for tomographic measurements on nuclear fuel are discussed. It is argued that the use of a detector system with high energy resolution and high peak efficiency in connection to spectroscopic peak analysis is beneficial.
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| 37. |
- 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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| 38. |
- 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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| 39. |
- Jansson, Peter, 1971-, et al.
(författare)
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A Feasibility Study of BGO Scintillation Detectors for Tomographic Measurements on Nuclear Fuel
- 2000
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- A study of BGO detectors has been performed. The purpose of the study was to determine geometrical shape of the scintillator crystals in order to be suited for use in tomographic measurements on nuclear fuel. Computer calculations using Monte Carlo techniques were used. High count-rate experiments were performed on three nuclear fuel assemblies with the shapes of the crystals determined by the calculations. The resulting characteristics of the detectors show that they are suitable in a tomographic measurement.
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| 40. |
- 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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| 41. |
- Jansson, Peter, 1971-, et al.
(författare)
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A Method of Measuring Decay Heat in Spent Nuclear Fuel using Gamma-ray Spectroscopy
- 2001
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Ingår i: Waste Management Symposium 2001 (WM'01).
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Konferensbidrag (refereegranskat)abstract
- In this paper, a method is presented for determining the decay heat in spent nuclear fuel by using gamma-ray spectroscopy. Using this method, the decay heat may be determined within ten minutes per assembly i.e. it is well suited for industrial applications in, for example, an encapsulation facility. The method has been tested and evaluated in the wet Swedish Central Storage for Spent Fuel, CLAB. Although only tested in a wet storage, the method should also be applicable for dry storage.The objective of developing the method was primarily to investigate possibilities to achieve a fast, robust and reasonable accurate determination of decay heat by gamma-ray measurements on fuel assemblies. Such a method could also be for verification of burnup and cooling time, for safeguard purposes prior to encapsulation, (1).So far, measurements and calculations on 35 BWR- and 34 PWR-assemblies, with various nuclear data, have been performed. The test measurements, using preliminary measuring equipment, have shown that the decay heat may be determined within an uncertainty of 3%.
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| 42. |
- Jansson, Peter, 1971-, et al.
(författare)
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Calculations of the Neutron Flux Outside BWR 8×8 Spent-Fuel Assemblies and the Sensitivity to Fuel Pin Diversion
- 2004
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Ingår i: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 146:1, s. 58-64
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Tidskriftsartikel (refereegranskat)abstract
- The possibility of detecting replaced fuel rods in a spent-fuel assembly by means of measurement of the emitted neutron- and gamma-ray radiation has been investigated by computer simulations. The radiation field outside a boiling water reactor 8 × 8 fuel assembly with varying patterns of fuel rods replaced with lead dummies was calculated using a simple model for the source distribution and the Monte Carlo code MCNP-4C for the radiation field. In particular, the sensitivity of the thermal neutron field as measured in a Fork detector to various replacement patterns was investigated. The results suggest a detection limit of 5% of the fuel mass replaced, i.e., 3 out of 63 rods, independently of the pattern of the replaced rods.
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| 43. |
- Jansson, Peter, 1971-, et al.
(författare)
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Gamma-ray measurements of spent PWR fuel and determination of residual power
- 1997
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The method for determining residual thermal power in spent BWR fuel described in ISV-4/97 have been used in an extended study where spent PWR fuel assemblies have been considered. The experimental work has been carried out at the interim storage CLAB. By using the 137Cs radiation it is shown in the present study that it is possible to experimentally determine the residual thermal power within 3%.
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| 44. |
- Jansson, Peter, 1971-, et al.
(författare)
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Gamma-Ray Spectroscopy Measurements of Decay Heat in Spent Nuclear Fuel
- 2002
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Ingår i: Nuclear Science and Engineering. - 0029-5639. ; 141:2, s. 129-139
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Tidskriftsartikel (refereegranskat)abstract
- A method for determining the residual thermal power in spent nuclear fuel using gamma-ray spectroscopy is suggested. It is based on the correlation between the residual power and the 137Cs activity, which is nearly linear for fuel with cooling times between 10 and 50 yr. Using available data of calorimetrically measured values of the decay heat in 69 boiling water reactor and pressurized water reactor spent-fuel assemblies resulted in agreement with a standard deviation of 3%.
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| 45. |
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| 46. |
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| 47. |
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| 48. |
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| 49. |
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| 50. |
- Lindberg, Bo, et al.
(författare)
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Modeling of Cherenkov light emission from BWR nuclear fuel with missing or substituted rods
- 2006
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Ingår i: IAEA Symposium on International Safeguards: Addressing Verification Challenges.
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Konferensbidrag (populärvet., debatt m.m.)abstract
- Computer simulations of Cherenkov glow from spent nuclear fuel were carried out. Spent nuclear fuel in storage ponds are verified with the help of the Cherenkov viewing device (CVD) and the Digital Cherenkov viewing device (DCVD). The instruments image the Cherenkov glow generated by gamma ray emissions from spent fuel into the water. An attempt to build a realistic digital model of the DCVD image containing partial-length, missing, and substituted rods was made to see if the effects of the deviations from normal can be predicted. It was concluded that partial-length or missing rods in the model was in good agreement with measured data, but replaced rods in the model showed a weaker attenuation of the Cherenkov glow than the observed DCVD images.
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