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BETA
Bäcklin, Anders
Alternative names
Publications (10 of 24) Show all publications
Osifo, O., Jacobsson Svärd, S., Håkansson, A., Willman, C., Bäcklin, A. & Lundqvist, T. (2008). Verification and determination of the decay heat in spent PWR fuel by means of gamma scanning. Nuclear science and engineering, 160(1), 129-143
Open this publication in new window or tab >>Verification and determination of the decay heat in spent PWR fuel by means of gamma scanning
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2008 (English)In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 160, no 1, p. 129-143Article in journal (Refereed) Published
Abstract [en]

Decay heat is an important design parameter at the future Swedish spent nuclear fuel repository. It will be calculated for each fuel assembly using dedicated depletion codes, based on the operator-declared irradiation history. However, experimental verification of the calculated decay heat is also anticipated. Such verification may, be obtained by, gamma scanning using the established correlation between the decay heat and the emitted gamma-ray intensity from Cs-137. In this procedure, the correctness of the operator-declared fuel parameters can be verified. Recent achievements of the gamma-scanning technique include the development of a dedicated spectroscopic data-acquisition system and the use of an advanced calorimeter for calibration. Using this system, the operator-declared burnup and cooling time of 31 pressurized water reactor fuel assemblies was verified experimentally, to within 2.2% (1 sigma) and 1.9% (1 sigma), respectively. The measured decay heat agreed with calorimetric data within 2.3% (1 sigma). whereby the calculated decay, heat was verified within 2.3% (1 sigma). The measuring time per fuel assembly was similar to 15 min. In case reliable operator-declared data are not available, the gamma-scanning technique also provides a means to independently measure the decay, heat. The results obtained in this procedure agreed with calorimetric data within 2.7% (1 sigma).

National Category
Physical Sciences
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-94791 (URN)000258579200009 ()
Available from: 2006-09-08 Created: 2006-09-08 Last updated: 2017-12-14Bibliographically approved
Jansson, P., Jacobsson Svärd, S., Håkansson, A. & Bäcklin, A. (2006). A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly. Nuclear Science and Engineering, 152(1), 76-86
Open this publication in new window or tab >>A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly
2006 (English)In: Nuclear Science and Engineering, ISSN 0029-5639, Vol. 152, no 1, p. 76-86Article in journal (Refereed) Published
Abstract [en]

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 sigma). It is primarily planned to be used for validation of production codes for core simulation but may also be used for safeguards purposes.

National Category
Subatomic Physics
Identifiers
urn:nbn:se:uu:diva-77336 (URN)
Available from: 2006-03-15 Created: 2006-03-15 Last updated: 2017-11-24
Jansson, P., Jacobsson Svärd, S., Håkansson, A. & Bäcklin, A. (2006). A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly. Nuclear science and engineering, 152(1), 76-86
Open this publication in new window or tab >>A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly
2006 (English)In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 152, no 1, p. 76-86Article in journal (Refereed) Published
Abstract [en]

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.

Keyword
tomography, nuclear fuel
National Category
Subatomic Physics
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-91655 (URN)
Available from: 2004-04-21 Created: 2004-04-21 Last updated: 2017-12-14
Willman, C., Håkansson, A., Osifo, O., Bäcklin, A. & Jacobsson Svärd, S. (2006). A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes. Annals of Nuclear Energy, 33(9), 766-773
Open this publication in new window or tab >>A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes
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2006 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 33, no 9, p. 766-773Article in journal (Refereed) Published
Abstract [en]

Plutonium-rich mixed oxide fuel (MOX) is increasingly used in thermal reactors. However, spent MOX fuel could be a potential source of nuclear weapons material and a safeguards issue is therefore to determine whether a spent nuclear fuel assembly is of MOX type or of LEU (Low Enriched Uranium) type.

In this paper, we present theoretical and experimental results of a study that aims to investigate the possibilities of using gamma-ray spectroscopy to determine whether a nuclear fuel assembly is of MOX or of LEU type.

Simulations with the computer code ORIGEN-ARP have been performed where LEU and MOX fuel types with varying enrichment and burnup as well as different irradiation histories have been modelled. The simulations indicate that the fuel type determination may be achieved by using the intensity ratio Cs-134/Eu-154.

An experimental study of MOX fuel of 14 x 14 PWR type and LEU fuel of both 15 x 15 and 17 x 17 type is also reported in this paper. The outcome of the experimental study support the conclusion that MOX fuel may be discriminated from LEU fuel by measuring the suggested isotopic ratio.

National Category
Physical Sciences
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-94789 (URN)10.1016/j.anucene.2006.04.006 (DOI)000239532900003 ()
Available from: 2006-09-08 Created: 2006-09-08 Last updated: 2017-12-14
Willman, C., Håkansson, A., Osifo, O., Bäcklin, A. & Jacobsson Svärd, S. (2006). Nondestructive assay of spent nuclear fuel with gamma-ray spectroscopy. Annals of Nuclear Energy, 33(5), 427-438
Open this publication in new window or tab >>Nondestructive assay of spent nuclear fuel with gamma-ray spectroscopy
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2006 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, Vol. 33, no 5, p. 427-438Article in journal (Refereed) Published
Abstract [en]

An important issue in nuclear safeguards is to verify operator-declared data of spent nuclear fuel. Various techniques have therefore been assigned for this purpose. A nondestructive approach is to measure the gamma radiation from spent nuclear fuel assemblies. Using this technique, parameters such as burnup and cooling time can be calculated or verified.

In this paper, we propose the utilization of gamma rays from 137Cs, 134Cs and 154Eu to determine the consistency of operator-declared information. Specifically, we have investigated to what extent irradiation histories can be verified.

Computer simulations were used in order to determine limits for detecting small deviations from declared data. In addition, the technique has been experimentally demonstrated on 12 PWR fuel assemblies.

A technique for determining burnup and cooling time for fuel assemblies where no operator-declared information is available is also presented. In such a case, the burnup could be determined with 1.6% relative standard deviation and the cooling time with 1.5%.

National Category
Subatomic Physics
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-77666 (URN)
Available from: 2007-02-13 Created: 2007-02-13 Last updated: 2012-03-09
Jacobsson Svärd, S., Håkansson, A., Bäcklin, A., Osifo, O., Willman, C. & Jansson, P. (2005). Nondestructive Experimental Determination of the Pin-Power Distribution in Nuclear Fuel Assemblies. Nuclear Technology, 151(1), 70-76
Open this publication in new window or tab >>Nondestructive Experimental Determination of the Pin-Power Distribution in Nuclear Fuel Assemblies
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2005 (English)In: Nuclear Technology, ISSN 0029-5450, Vol. 151, no 1, p. 70-76Article in journal (Refereed) Published
Abstract [en]

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.

Keyword
tomography
National Category
Subatomic Physics
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-77332 (URN)
Available from: 2006-03-15 Created: 2006-03-15 Last updated: 2017-01-21
Jansson, P., Håkansson, A. & Bäcklin, A. (2004). Calculations of the Neutron Flux Outside BWR 8×8 Spent-Fuel Assemblies and the Sensitivity to Fuel Pin Diversion. Nuclear Technology, 146(1), 58-64
Open this publication in new window or tab >>Calculations of the Neutron Flux Outside BWR 8×8 Spent-Fuel Assemblies and the Sensitivity to Fuel Pin Diversion
2004 (English)In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 146, no 1, p. 58-64Article in journal (Refereed) Published
Abstract [en]

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.

Keyword
FORK, neutron, nuclear fuel, diversion, safeguards
National Category
Subatomic Physics
Research subject
Applied Nuclear Physics; Physics with specialization in Applied Nuclear Physics
Identifiers
urn:nbn:se:uu:diva-77331 (URN)10.13182/NT04-A3487 (DOI)
Available from: 2006-03-15 Created: 2006-03-15 Last updated: 2018-02-05
Jacobsson Svärd, S., Håkansson, A., Bäcklin, A., Osifo, O., Willman, C. & Jansson, P. (2003). Non-destructive experimental determination of the pin-power distribution in nuclear fuel. In: : . Paper presented at Advances in Nuclear Fuel Management III.
Open this publication in new window or tab >>Non-destructive experimental determination of the pin-power distribution in nuclear fuel
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2003 (English)Conference paper, Published paper (Other academic)
Abstract [en]

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.

National Category
Subatomic Physics
Identifiers
urn:nbn:se:uu:diva-77504 (URN)0-89448--670-5 (ISBN)
Conference
Advances in Nuclear Fuel Management III
Available from: 2006-03-15 Created: 2006-03-15 Last updated: 2015-01-27
Håkansson, A., Jacobsson Svärd, S. & Bäcklin, A. (2003). Vad gammastrålning kan berätta om kärnbränsle. KOSMOS , Årsbok för Svenska Fysikersamfundet
Open this publication in new window or tab >>Vad gammastrålning kan berätta om kärnbränsle
2003 (Swedish)In: KOSMOS , Årsbok för Svenska FysikersamfundetArticle in journal (Other (popular scientific, debate etc.)) Published
Identifiers
urn:nbn:se:uu:diva-26081 (URN)
Available from: 2007-02-14 Created: 2007-02-14 Last updated: 2011-01-13
Jansson, P., Håkansson, A. & Bäcklin, A. (2002). Detection of Partial Defects in Irradiated BWR Fuel Assemblies. A Preliminary Study. In: INMM 43rd Annual Meeting (INMM),Orlando, Florida, USA, June 23-27, 2002.
Open this publication in new window or tab >>Detection of Partial Defects in Irradiated BWR Fuel Assemblies. A Preliminary Study
2002 (English)In: INMM 43rd Annual Meeting (INMM),Orlando, Florida, USA, June 23-27, 2002, 2002Conference paper, Published paper (Refereed)
Identifiers
urn:nbn:se:uu:diva-25549 (URN)
Available from: 2007-02-13 Created: 2007-02-13
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