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Dalla Betta, G.-F., Obryk, B., Pia, M. G., Britton, C., Cao, L. R., Dong, Z., . . . Lyoussi, A. (2020). Comments by the Senior Editor. Paper presented at Advancements in Nuclear Instrumentation Measurement Methods and their Applications (ANIMMA). IEEE Transactions on Nuclear Science, 67(4), 543-543
Open this publication in new window or tab >>Comments by the Senior Editor
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2020 (English)In: IEEE Transactions on Nuclear Science, ISSN 0018-9499, E-ISSN 1558-1578, Vol. 67, no 4, p. 543-543Article in journal, Editorial material (Refereed) Published
Abstract [en]

The sixth edition of the International Conference on “Advancements in Nuclear Instrumentation Measurement Methods and their Applications” (ANIMMA) was held in Portorož, Slovenia, from June 17–21, 2019. The conference attracted almost 300 participants from 32 different countries, coming from academy, research institutes, and industry to discuss new scientific and technical prospects in all fields where nuclear instrumentation and measurements techniques play a major role. The scientific program included 25 invited talks, 125 oral presentations, and 77 poster presentations on the following topics: fundamental physics; fusion diagnostics and technology; nuclear power reactors monitoring and control; research reactors; nuclear fuel cycle; decommissioning, dismantling, and remote handling; safeguards and homeland security; severe accident monitoring; environmental and medical sciences; and education, training, and outreach.

National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-410727 (URN)10.1109/TNS.2020.2984961 (DOI)
Conference
Advancements in Nuclear Instrumentation Measurement Methods and their Applications (ANIMMA)
Available from: 2020-05-18 Created: 2020-05-18 Last updated: 2020-05-18
Jansson, P., Bengtsson, M., Bäckström, U., Svensson, K., Lycksell, M. & Sjöland, A. (2020). Data from calorimetric decay heat measurements of five used PWR 17x17 nuclear fuel assemblies. Data in Brief, 28, Article ID 104917.
Open this publication in new window or tab >>Data from calorimetric decay heat measurements of five used PWR 17x17 nuclear fuel assemblies
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2020 (English)In: Data in Brief, E-ISSN 2352-3409, Vol. 28, article id 104917Article in journal (Refereed) Published
Abstract [en]

Raw data from calorimetric measurements of five nuclear fuel assemblies of the PWR 17x17 type are provided. Measurements of the temperature both inside a calorimeter, in which the fuel assembly was placed, as well as outside, were performed as a function of time while water circulating inside the calorimeter heats up from radiation emitted in the radioactive decay of material in the fuel assembly. The data contain also measurements of dose rate in the water outside the calorimeter. Data from 38 measurements using an electrically heated model of a fuel assembly are also provided to be used for, e.g.¸ calibration.

The data can be used for validation of computer codes used for modelling of nuclear systems used for, e.g. nuclear reactors, storage and transport of nuclear fuel or systems for geological disposal.

Keywords
Spent nuclear fuel, decay heat, calorimetry, calorimetric measurement
National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-398177 (URN)10.1016/j.dib.2019.104917 (DOI)000520402100109 ()31890784 (PubMedID)
Funder
Swedish Nuclear Fuel and Waste Management Company, SKB, 21969
Available from: 2019-12-03 Created: 2019-12-03 Last updated: 2020-05-06Bibliographically approved
Andersson, P., Rathore, V., Senis, L., Anastasiadis, A., Andersson Sundén, E., Atak, H., . . . Nyberg, J. (2020). Simulation of the response of a segmented High-Purity Germanium detector for gamma emission tomography of nuclear fuel. SN Applied Sciences, 2, Article ID 271.
Open this publication in new window or tab >>Simulation of the response of a segmented High-Purity Germanium detector for gamma emission tomography of nuclear fuel
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2020 (English)In: SN Applied Sciences, ISSN 2523-3963, Vol. 2, article id 271Article in journal (Refereed) Published
Abstract [en]

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.

Place, publisher, year, edition, pages
Springer, 2020
National Category
Engineering and Technology Accelerator Physics and Instrumentation
Identifiers
urn:nbn:se:uu:diva-392188 (URN)10.1007/s42452-020-2053-4 (DOI)000517964300132 ()
Funder
Swedish Research Council, 2017-06448Swedish Foundation for Strategic Research , EM16-0031
Available from: 2019-08-30 Created: 2019-08-30 Last updated: 2020-04-02Bibliographically approved
Atak, H., Anastasiadis, A., Jansson, P., Elter, Z., Andersson Sundén, E., Holcombe, S. & Andersson, P. (2020). The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup. Progress in nuclear energy (New series), 125, Article ID 103359.
Open this publication in new window or tab >>The degradation of gamma-ray mass attenuation of UOX and MOX fuel with nuclear burnup
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2020 (English)In: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 125, article id 103359Article in journal (Refereed) Published
Abstract [en]

Nondestructive gamma-ray spectrometry of nuclear fuel is routinely performed in axial gamma scanning devices and more recently with gamma emission tomography. Following the irradiation of a fresh nuclear fuel with high intensity neutron flux in a nuclear reactor core, a great number of gamma-emitting radionuclides are created. These can be utilized for gamma spectrometric techniques. However, due to the high density and atomic number of the nuclear fuel, self-attenuation of gamma-rays is a challenge, which requires attenuation correction in order to perform accurate analysis of the source activity in the fuel.

In this study, the degradation of the gamma-ray mass attenuation with burnup was investigated and, in addition, a predictive model was created by investigating the attenuation change at various gamma energies caused by the burnup of the nuclear fuel. This model is intended for use by spectrometry practitioners inspecting nuclear fuel. To this aim, the energy-dependent gamma-ray mass-attenuation coefficients were investigated as a function of burnup for UOX, and three MOX fuels having different initial Pu contents. The Serpent 2 reactor physics code was used to simulate the burnup history of the fuel pins. The nuclide inventory of the Serpent 2 output is combined with the NIST XCOM database to calculate the mass attenuation coefficients.

The mass attenuation coefficient of the fuel was found to decrease with the fuel burnup, in the range of a few percent, depending on the burnup and gamma energy. Also, a theoretical burnup dependent swelling model was imposed on fuel density to see how linear attenuation coefficient of fuel material is changed. Furthermore, greater effect may be expected on the transmitted intensity, where a simulation study of a PWR assembly revealed that the contribution from the inner rods in a scanned fuel assembly increased by tens of percent compared to the one with non-irradiated fresh fuels, when shielded by the outer rods of the assembly. A sensitivity analysis was performed in order to test the effect of a number of geometrical and operational reactor parameters that were considered to potentially effect the mass attenuation coefficient. Finally, a simple-to-use predictive model was constructed providing the mass-attenuation coefficient [cm2/g] of fuel as a function of burnup [MWd/kgHM] and initial Pu content [wt%]. The resulting predictive model was optimized by using the nonlinear regression method.

Keywords
Gamma-ray spectrometry, Emission tomography, Nuclear fuel assembly, Nondestructive measurements, High burnup, Gamma-ray self-attenuation
National Category
Energy Engineering
Identifiers
urn:nbn:se:uu:diva-395145 (URN)10.1016/j.pnucene.2020.103359 (DOI)
Funder
Swedish Research Council, 2017-06448
Available from: 2019-10-14 Created: 2019-10-14 Last updated: 2020-04-28Bibliographically approved
Jansson, P., Schillebeeckx, P., Zencker, U. & Cobos, J. (2019). EURAD: SFC WP: Spent Fuel Characterisation and Evolution Until Disposal. In: : . Paper presented at Euradwaste ’19, 4-7 June 2019, Pitesti, Romania.
Open this publication in new window or tab >>EURAD: SFC WP: Spent Fuel Characterisation and Evolution Until Disposal
2019 (English)Conference paper, Poster (with or without abstract) (Other academic)
Keywords
nuclear fuel, characterization, back-end
National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-389524 (URN)
Conference
Euradwaste ’19, 4-7 June 2019, Pitesti, Romania
Available from: 2019-07-17 Created: 2019-07-17 Last updated: 2019-08-09Bibliographically approved
Jansson, P. (2019). Geant4 Monte Carlo calculations of gamma- and neutron flux and energy deposition in water outside the PWR 17x17 nuclear fuel assembly BT01.
Open this publication in new window or tab >>Geant4 Monte Carlo calculations of gamma- and neutron flux and energy deposition in water outside the PWR 17x17 nuclear fuel assembly BT01
2019 (English)Data set, Primary data
Abstract [en]

Using the Geant4 Monte Carlo framework, the flux of radiation emitted from the used nuclear fuel assembly BT01 has been estimated. The emission energy spectra for photons and neutrons calculated using SCALE for this PWR 17x17 fuel assembly were used as source for the radiation transport. The fuel assembly was modelled with the fuel rods emersed in a universe comprised of water. The photon and neutron flux and energy deposition from the source in a plane at the centre of the fuel assembly were calculated.

This dataset contains the raw data files produced by the calculations.

National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-407856 (URN)
Available from: 2020-03-30 Created: 2020-03-30 Last updated: 2020-04-03Bibliographically approved
Elter, Z., Mishra, V., Grape, S., Branger, E., Jansson, P. & Caldeira Balkeståhl, L. (2019). Investigating the gamma and neutron radiation around quivers for verification purposes. In: : . Paper presented at 41st ESARDA Annual Meeting Symposium, 14-16 May, 2019,Stresa, Italy.
Open this publication in new window or tab >>Investigating the gamma and neutron radiation around quivers for verification purposes
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2019 (English)Conference paper, Oral presentation with published abstract (Other academic)
Abstract [en]

Before encapsulation of spent nuclear fuel in a geological repository, the fuels need to be verified fors afeguards purposes. This requirement applies to all spent fuel assemblies, including those with properties or designs that are especially challenging to verify. One such example are quivers, a new type of containers used to hold damaged spent fuel rods. After placing damaged rods inside the quivers, they are sealed with a thick lid and the water is removed. The lid is thick enough to significantly reduce the amount of the gamma radiation penetrating through it, which can make safeguards verification from the top using gamma techniques difficult.

In this paper we make a first feasibility study related to safeguards verification of quivers, aimed at investigating the gamma and neutron radiation field around a quiver using a simplified quiver geometry. The nuclide inventory of the rods placed in the quiver is calculated with Serpent and Origen-Arp, and the radiation transport is modeled with Serpent. The objective is to assess the capability of existing non-destructive assay instruments, measuring the gamma and/or neutron radiation from the object, to verify the content for nuclear safeguards purposes. The results show that the thick quiver lid attenuates the gamma radiation, thereby making gamma-radiation based verification from above the quiver difficult. Verification using neutron instruments above the quiver, or gamma and/or neutron instruments on the side may be possible. These results are in agreement with measurements of a BWR quiver using a DCVD, performed by the authors.

Keywords
quiver, safeguards verification, gamma radiation, neutron radiation, spent fuel, PWR, modeling
National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-389765 (URN)
Conference
41st ESARDA Annual Meeting Symposium, 14-16 May, 2019,Stresa, Italy
Available from: 2019-07-24 Created: 2019-07-24 Last updated: 2020-05-06Bibliographically approved
Grape, S., Branger, E., Elter, Z., Jansson, P. & Mishra, V. (2019). Machine learning in nuclear safeguards. In: : . Paper presented at Swedish Workshop on Data Science.
Open this publication in new window or tab >>Machine learning in nuclear safeguards
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2019 (English)Conference paper, Poster (with or without abstract) (Other academic)
Abstract [en]

•Before placing spent nuclear fuel in in a geological repository, they will be characterized and their declared properties will be verified.

•We have created large library of modelled spent nuclear fuel (SNF) assemblies and estimated their activity of gamma-ray emitting fission products, the early die-away time τ and the Cherenkov light intensity.

•We have used Random Forest regression to evaluate the capability to determine the fuel parameters initial enrichment (IE), burnup (BU) and cooling time (CT) using data from non-destructive assay (NDA) techniques

National Category
Other Physics Topics
Identifiers
urn:nbn:se:uu:diva-394881 (URN)
Conference
Swedish Workshop on Data Science
Funder
Swedish Radiation Safety Authority
Available from: 2019-10-10 Created: 2019-10-10 Last updated: 2019-10-10
Branger, E., Grape, S., Jansson, P. & Jacobsson Svärd, S. (2019). On the inclusion of light transport in prediction tools for Cherenkov light intensity assessment of irradiated nuclear fuel assemblies. Journal of Instrumentation, 14, Article ID T01010.
Open this publication in new window or tab >>On the inclusion of light transport in prediction tools for Cherenkov light intensity assessment of irradiated nuclear fuel assemblies
2019 (English)In: Journal of Instrumentation, ISSN 1748-0221, E-ISSN 1748-0221, Vol. 14, article id T01010Article in journal (Refereed) Published
Abstract [en]

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.

Keywords
Nuclear safeguards, Geant4, Cherenkov light, DCVD, Nuclear fuel
National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-357151 (URN)10.1088/1748-0221/14/01/T01010 (DOI)000457930800001 ()
Funder
Swedish Radiation Safety Authority, SSM2012-2750Swedish National Infrastructure for Computing (SNIC), p2007011
Available from: 2018-08-13 Created: 2018-08-13 Last updated: 2019-03-05Bibliographically approved
Janssens, W., Niemeyer, I., Aregbe, Y., Bonino, F., Funk, P., Hildingsson, L., . . . Vince, A. (2019). Outcome and Actions of the 2019 Reflection Group of the European Safeguards Research and Development Association (ESARDA). In: : . Paper presented at INMM 60:th Annual Meeting.
Open this publication in new window or tab >>Outcome and Actions of the 2019 Reflection Group of the European Safeguards Research and Development Association (ESARDA)
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2019 (English)Conference paper, Published paper (Other academic)
Abstract [en]

The European Safeguards Research and Development Association (ESARDA), founded in 1969, is a voluntary association of European organizations formed to foster, advance and harmonize research and development (R&D) in the area of nuclear safeguards. It provides a forum for the exchange of information and ideas between nuclear facility operators, safeguards national authorities, regional and international inspectorates, and individuals engaged in safeguards-related research and development. Today ESARDA includes 33 Parties from within the European Union. In addition, a further eight laboratories, authorities, operators and academic institutions from outside the EU have joined ESARDA as Associate Members, while the Association signed Memoranda of Understanding with the Asia-Pacific Safeguards Network and the African Commission on Nuclear Energy, and a Letter of Intent with the Institute for Nuclear Materials Management.

ESARDA seeks to maintain a dynamic approach to the developing priorities, while ensuring that its activities continue to anticipate future needs, which is why the Association periodically undertakes a formal Reflection Group exercise. In the last 2 years, the Reflection Group, RG2019, worked along the following objectives:

  1. develop a roadmap to improve and enhance the quality, effectiveness and efficiency of safeguards and non-proliferation in Europe and abroad; and
  2. ensure that the future activities of ESARDA are both consistent with the Association’s purpose, as stated in the ESARDA Agreement, and address the needs of the ESARDA members and/or stakeholders.

In the report, finalized before the ESARDA Symposium in May 2019, three specific goals were identified:

  1. establish short term ESARDA priorities (2019 to 2024) and prepare a roadmap - i.e. WHAT;
  2. define ESARDA’s long-term future (2019-2050) activities based on the new landscape in Europe and internationally - to be reviewed before 2025 to establish the next 5 year plan; and
  3. review the ESARDA organization, and discuss HOW ESARDA can achieve the identified objectives and implement the identified roadmap.

A World Café on these topics was organized and held during the 2019 Symposium. In this paper, the key outcomes and results of the ESARDA Reflection Group 2019 are presented, including their relevance for the international partners of ESARDA.

National Category
Subatomic Physics
Research subject
Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-392277 (URN)
Conference
INMM 60:th Annual Meeting
Available from: 2019-09-02 Created: 2019-09-02 Last updated: 2019-09-02
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-3136-5665

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