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  • 1.
    Alhassan, Erwin
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Sjöstrand, Henrik
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Helgesson, Petter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Arjan, J. Koning
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Pomp, Stephan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Dimitri, Rochman
    Nuclear Research and Consultancy Group.
    Uncertainty and correlation analysis of lead nuclear data on reactor parameters for the European Lead Cooled Training Reactor2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 75, p. 26-37Article in journal (Refereed)
    Abstract [en]

    The Total Monte Carlo (TMC) method was used in this study to assess the impact of Pb-204, 206, 207, 208 nuclear data uncertainties on reactor safety parameters for the ELECTRA reactor. Relatively large uncertainties were observed in the k-eff and the coolant void worth (CVW) for all isotopes except for Pb-204 with signicant contribution coming from Pb-208 nuclear data; the dominant eectcame from uncertainties in the resonance parameters; however, elastic scattering cross section and the angular distributions also had signicant impact. It was also observed that the k-eff distribution for Pb-206, 207, 208 deviates from a Gaussian distribution with tails in the high k-eff region. An uncertainty of 0.9% on the k-eff and 3.3% for the CVW due to lead nuclear data were obtained. As part of the work, cross section-reactor parameter correlations were also studied using a Monte Carlo sensitivity method. Strong correlations were observed between the k-eff and (n,el) cross section for all the lead isotopes. The correlation between the (n,inl) and the k-eff was also found to be signicant.

  • 2.
    Alhassan, Erwin
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Sjöstrand, Henrik
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Helgesson, Petter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Pomp, Stephan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Arjan, J. Koning
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. Int Atom Energy Commiss, Nucl Data Sect, Vienna, Austria.
    Rochman, Dimitri
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, CH-5232 Villigen, Switzerland.
    Selecting benchmarks for reactor simulations: an application to a Lead Fast Reactor2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 96, p. 158-169Article in journal (Refereed)
    Abstract [en]

    For several decades reactor design has been supported by computer codes for the investigation of reactor behavior under both steady state and transient conditions. The use of computer codes to simulate reactor behavior enables the investigation of various safety scenarios saving time and cost. There has been an increase in the development of in-house (local) codes by various research groups in recent times for preliminary design of specific or targeted nuclear reactor applications. These codes must be validated and calibrated against experimental benchmark data with their evolution and improvements. Given the large number of benchmarks available, selecting these benchmarks for reactor calculations and validation of simulation codes for specific or target applications can be rather tedious and difficult. In the past, the traditional approach based on expert judgement using information provided in various handbooks, has been used for the selection of these benchmarks. This approach has been criticized because it introduces a user bias into the selection process. This paper presents a method for selecting these benchmarks for reactor calculations for specific reactor applications based on the Total Monte Carlo (TMC) method. First, nuclear model parameters are randomly sampled within a given probability distribution and a large set of random nuclear data files are produced using the TALYS code system. These files are processed and used to analyze a target reactor system and a set of criticality benchmarks. Similarity between the target reactor system and one or several benchmarks is quantified using a similarity index. The method has been applied to the European Lead Cooled Reactor (ELECTRA) and a set of plutonium and lead sensitive criticality benchmarks using the effective multiplication factor (keffkeff). From the study, strong similarity were observed in the keffkeff between ELECTRA and some plutonium and lead sensitive criticality benchmarks. Also, for validation purposes, simulation results for a list of selected criticality benchmarks simulated with the MCNPX and SERPENT codes using different nuclear data libraries have been compared with experimentally measured benchmark keff values.

  • 3.
    Alhassan, Erwin
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Sjöstrand, Henrik
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Helgesson, Petter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Pomp, Stephan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Koning, Arjan J.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. Nuclear Research and Consultancy Group, Petten, The Netherlands.
    Rochman, D.
    Laboratory for Reactor Physics Systems Behaviour, Paul Scherrer Institut, Villigen, Switzerland.
    Benchmark selection methodology for reactor calculations and nuclear data uncertainty reduction2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
    Abstract [en]

    Criticality, reactor physics and shielding benchmarks are expected to play important roles in GEN-IV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used for validating computer codes and for testing nuclear data libraries. Given the large number of benchmarks available, selecting these benchmarks for specic applications can be rather tedious and difficult. Until recently, the selection process has been based usually on expert judgement which is dependent on the expertise and the experience of the user and there by introducing a user bias into the process. This approach is also not suitable for the Total Monte Carlo methodology which lays strong emphasis on automation, reproducibility and quality assurance. In this paper a method for selecting these benchmarks for reactor calculation and for nuclear data uncertainty reduction based on the Total Monte Carlo (TMC) method is presented. For reactor code validation purposes, similarities between a real reactor application and one or several benchmarks are quantied using a similarity index while the Pearson correlation coecient is used to select benchmarks for nuclear data uncertainty reduction. Also, a correlation based sensitivity method is used to identify the sensitivity of benchmarks to particular nuclear reactions. Based on the benchmark selection methodology, two approaches are presented for reducing nuclear data uncertainty using integral benchmark experiments as an additional constraint in the TMC method: a binary accept/reject and a method of assigning file weights using the likelihood function. Finally, the methods are applied to a full lead-cooled fast reactor core and a set of criticality benchmarks. Signicant reductions in Pu-239 and Pb-208 nuclear data uncertainties were obtained after implementing the two methods with some benchmarks.

  • 4.
    Andersson, Peter
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Holcombe, Scott
    OECD Halden Reactor Project, Inst Energy Technol, Box 173, NO-1751 Halden, Norway.
    A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 110, p. 88-97Article in journal (Refereed)
    Abstract [en]

    The Halden reactor project (HRP) has recently developed a gamma emission tomography instrument dedicated for measurements of irradiated nuclear fuel in collaboration with Westinghouse and Uppsala University. This instrument is now assembled and the first experimental measurements have been performed on fuel assemblies irradiated in the Halden reactor. The objective of the instrument is to map the distribution of radioisotopes of interest in the fuel, e.g. 137Cs or 140La/Ba, and for this purpose, a spectroscopic high-purity Germanium detector has been selected, which enables the identification and tomographic reconstruction of separate isotopes by their characteristic gamma rays.

    To gain from the analysis of the data from this new instrument, and in the future from other gamma emission tomography instruments for nuclear fuels, various reconstruction methods are available that vary in the accuracy and the amount of detail obtainable in the analysis. This paper presents the details of the working principles of a new code for gamma emission tomography, the UPPREC (UPPsala university REConstruction) code. It is a development in MATLABTM code with the aim to produce detailed quantitative images of the investigated fuel.

    In this paper, the methods assembled for the analysis of data collected by this novel instrument are described and demonstrated and a benchmark is presented using single rod gamma scanning. It is shown that the UPPREC methodology improves the accuracy of the reconstructions by removing the errors introduced by the presence of highly attenuating fuel and structural material in the fuel assembly. With the introduction of UPPREC, detailed quantitative cross-sectional images of nuclide concentrations in nuclear fuel are now able to be obtained by nondestructive means. This has potential applications in both nuclear fuel diagnostics and in safeguards.

  • 5. Bansah, C. Y.
    et al.
    Akaho, E. H. K.
    Ayensu, A.
    Adoo, N. A.
    Agbodemegbe, V. Y.
    Alhassan, Erwin
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Della, R.
    Theoretical model for predicting the relative timings of potential failures in steam generator tubes of a PWR during a severe accident2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 59, p. 10-15Article, review/survey (Refereed)
    Abstract [en]

    During certain severe reactor accidents such as station-blackout accidents, countercurrent natural circulation flow could develop within the reactor coolant system. Natural circulation flow is very important because of transfer of decay energy from the core to other parts of the reactor coolant system. The associated heat-ups of the reactor coolant system structures can lead to pressure boundary failures with notable vulnerabilities in the pressurizer surge line, the hot leg nozzles and the steam generator tubes. The potential for a steam generator tube failure has been of particular concern because fission products could be released to the environment through such a failure. To solve the problem of steam generator tube failure, a computer code - Steam Generator Mitigation Program (SGMP), written in FORTRAN 95 computes the recirculation ratio (RR) and the mixing fraction (MF) which are the main parameters used in characterizing natural circulation. In the flow analysis, the RR and MF were respectively found to be 2.4 +/- 0.3 and 0.8 +/- 0.17. The results obtained showed that the natural circulation would delay the failure time of the steam generator tubes and is in good qualitative agreement with results from literature. 

  • 6.
    Boafo, E.K.
    et al.
    National Nuclear Research Institute, Ghana Atomic Energy Commission.
    Alhassan, Erwin
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Akaho, E.H.K.
    National Nuclear Research Institute, Ghana Atomic Energy Commission.
    Utilizing the burnup capability in MCNPX to perform depletion analysisof an MNSR fuel2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 73, p. 478-483Article in journal (Refereed)
    Abstract [en]

    In this work, we present results of fuel depletion analyses performed for a potential LEU core of Ghana’s Miniature Neutron Source Reactor (GHARR-1) using the Monte Carlo N-particle extended (MCNPX) neutron transport code. Depletion calculation was carried out for the reactor core from the Beginning of Life (BOL) to the End of Life (EOL) which corresponds to 10 years of reactor operation. The amounts of uranium and plutonium actinides were estimated at BOL and EOL of the core. Decay heat removal rate for the MNSR after reactor shut down was investigated due to its significance to reactor safety. Inventory of fission products produced as a result of burnup was also calculated. The results show that a maximum discharge burnup equivalent to 0.568% of U-235 was consumed at EOL equivalent to operating the reactor for 200 Effective Full Power Days (EFPD), while the amount of Pu-239 produced was not significant.Also, the decay heat decreased exponentially after reactor shutdown confirming that decay heat will be removed in the system by natural circulation after shutdown and thus guaranteeing the safety of the reactor.

  • 7. Chernitskiy, S. V.
    et al.
    Moiseenko, V. E.
    Noack, Klaus
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Ågren, Olov
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Abdullayev, A.
    Static neutronic calculation of a subcritical transmutation stellarator-mirror fusion-fission hybrid2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 72, p. 413-420Article in journal (Refereed)
    Abstract [en]

    The MCNPX Monte-Carlo code has been used to model the neutron transport in a sub-critical fast fission reactor driven by a fusion neutron source. A stellarator-mirror device is considered as the fusion neutron source. The principal composition for a fission blanket of a mirror fusion-fission hybrid is devised from the calculations. Heat load on the first wall, the distribution of the neutron fields in the reactor, the neutron spectrum and the distribution of energy release in the blanket are calculated. The possibility of tritium breeding inside the installation in quantities that meet the needs of the fusion neutron source is analyzed. The portion of the plasma column generates fusion neutrons that mainly do not reach the fission reactor core is proposed to be surrounded by a vessel filled with borated water to absorb the flying out neutrons. The flux of the neutrons escaping from the device to surrounding space is also calculated.

  • 8.
    Dahlfors, Marcus
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics.
    Kadi, Yacine
    Herrera-Martínez, Adonai
    Neutron Cross Section Sensitivity for Minor Actinide Transmutation in Energy Amplifier Systems2007In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 34, no 10, p. 824-835Article in journal (Refereed)
    Abstract [en]

    The nuclear data sensitivity in 3D Monte Carlo burnup calculations of minor actinide transmutation in Energy Amplifier Systems is assessed. Ansaldo Nucleare's 80 MWth, Energy Amplifier Demonstration Facility (EADF) design serves as a technical and geometrical platform for the analysis. The accelerator-d riven EADF is a fast, subcritical system based on classical MOX-fuel technology and on molten lead-bismuth eutectic cooling. For Monte Carlo simulations, the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN, is utilised. The code offers treatment of both high-energy particle interactions and low-energy neutron transport with a sophisticated method based on a full Monte Carlo simulation, together with the option of employing different modern nuclear data libraries. In particular, the impact of nuclear data discrepancies on transmutation properties prediction with increasing exposure is examined. Monte Carlo simulation accelerator-driven systems transmutation burnup fast neutron spectrum minor actinides nuclear waste.

  • 9.
    Davour, Anna
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Andersson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. OECD Halden Reactor Project, Halden, Norway.
    Grape, Sophie
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Holcombe, Scott
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. OECD Halden Reactor Project, Halden, Norway.
    Jansson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Troeng, Mats
    Applying image analysis techniques to tomographic images of irradiated nuclear fuel assemblies2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 96, p. 223-229Article in journal (Refereed)
    Abstract [en]

    In this paper we present a set of image analysis techniques used for extraction of information from cross-sectional images of nuclear fuel assemblies, achieved from gamma emission tomography measurements. These techniques are based on template matching, an established method for identifying objects with known properties in images.

    We demonstrate a rod template matching algorithm for identification and counting of the fuel rods present in the image. This technique may be applicable in nuclear safeguards inspections, because of the potential of verifying the presence of all fuel rods, or potentially discovering any that are missing.

    We also demonstrate the accurate determination of the position of a fuel assembly, or parts of the assembly, within the imaged area. Accurate knowledge of the assembly position enables detailed modelling of the gamma transport through the fuel, which in turn is needed to make tomographic reconstructions quantifying the activity in each fuel rod with high precision.

    Using the full gamma energy spectrum, details about the location of different gamma-emitting isotopes within the fuel assembly can be extracted. We also demonstrate the capability to determine the position of supporting parts of the nuclear fuel assembly through their attenuating effect on the gamma rays emitted from the fuel. Altogether this enhances the capabilities of non-destructive nuclear fuel characterization.

  • 10.
    Helgesson, Petter
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Sjöstrand, Henrik
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Treating model defects by fitting smoothly varying model parameters: Energy dependence in nuclear data evaluation2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 120, p. 35-47Article in journal (Refereed)
    Abstract [en]

    The fitting of models to data is essential in nuclear data evaluation, as in many other fields of science. The models maybe necessary for interpolation or extrapolation, but they are seldom perfect; there are model defects present which can result in severe biases and underestimated uncertainties. This work presents and investigates the idea to treat this problem by letting the model parameters vary smoothly with an input parameter. To be specific, the model parameters for neutron cross sections are allowed to vary with neutron energy. The parameter variation is limited by Gaussian processes, but the method should not be confused with adding a Gaussian process to the model. The performance of the method is studied using a large number of synthetic data sets, such that it is possible to quantitatively study the distribution of results compared to the underlying truth. There are imperfections in the results, but the method is seen to readily outperform fits without the energy dependent parameters. In particular, the estimates of uncertainty and correlations are much better. Hence, the method seems to offer a promising route for future treatment of model defects, both for nuclear data and elsewhere.

  • 11.
    Hellesen, C.
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Qvist, S.
    Uppsala Univ, Uppsala, Sweden.;Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Benchmark and demonstration of the CHD code for transient analysis of fast reactor systems2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 109, p. 712-719Article in journal (Refereed)
    Abstract [en]

    In this paper the dynamic thermal hydraulic fast reactor simulation code CHD is presented. The code is built around a scriptable object-oriented framework in the programming language Python to be able to flexibly describe different reactor geometries including thermal-hydraulics models of an arbitrary number of coolant channels as well as pumps, heat-exchangers and pools etc. In addition, custom objects such as the Autonomous Reactivity Control (ARC) system for enhanced passive safety are modeled in detail. In this paper we compare the performance of the CHD code with other similar fast reactor dynamics codes using a benchmark study of the European Sodium cooled Fast Reactor (ESFR). The results agree well, both qualitatively and quantitatively with the code benchmark. In addition, we demonstrate the code's ability to simulate the long-term asymptotic behavior of a neutronically shut down reactor in an unprotected loss of flow scenario using a model of the Advanced Burner Reactor (ABR). (C) 2017 Elsevier Ltd. All rights reserved.

  • 12.
    Hellesen, Carl
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Grape, Sophie
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jansson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Åberg Lindell, Matilda
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Andersson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Nuclear Spent Fuel Parameter Determination using Multivariate Analysis of Fission Product Gamma Spectra2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 110, p. 886-895Article in journal (Refereed)
    Abstract [en]

    In this paper, we investigate the application of multivariate data analysis methods to the analysis of gamma spectroscopy measurements of spent nuclear fuel (SNF). Using a simulated irradiation and cooling of nuclear fuel over a wide range of cooling times (CT), total burnup at discharge (BU) and initial enrichments (IE) we investigate the possibilities of using a multivariate data analysis of the gamma ray emission signatures from the fuel to determine these fuel parameters. This is accomplished by training a multivariate analysis method on simulated data and then applying the method to simulated, but perturbed, data.

    We find that for SNF with CT less than about 20 years, a single gamma spectrum from a high resolution gamma spectrometer, such as a high-purity germanium spectrometer, allows for the determination of the above mentioned fuel parameters.

    Further, using measured gamma spectra from real SNF from Swedish pressurized light water reactors we were able to confirm the operator declared fuel parameters. In this case, a multivariate analysis trained on simulated data and applied to real data was used.

  • 13.
    Holcombe, Scott
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Eitrheim, Knut
    Hallstadius, Lars
    Willman, Christofer
    Feasibility of identifying leaking fuel rods using gamma tomography2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 57, p. 334-340Article in journal (Refereed)
    Abstract [en]

    In cases of fuel failure in irradiated nuclear fuel assemblies, causing leakage of fission gasses from a fuel rod, there is a need for reliable non-destructive measurement methods that can determine which rod is failed. Methods currently in use include visual inspection, eddy current, and ultrasonic testing, but additional alternatives have been under consideration, including tomographic gamma measurements.

    The simulations covered in this report show that tomographic measurements could be feasible. By measuring a characteristic gamma energy from fission gasses in the gas plenum, the rod-by-rod gamma source distribution within the fuel rod plena may be reconstructed into an image or data set which could then be compared to the predicted distribution of fission gasses, e.g. from the STAV code. Rods with significantly less fission gas in the plenum may then be identified as leakers.

    Results for rods with low fission gas release may, however, in some cases be inconclusive since these rods will already have a weak contribution to the measured gamma-ray intensities and for such rods there is a risk that a further decrease in fission gas content due to a leak may not be detectable. In order to evaluate this and similar experimental issues, measurement campaigns are planned using a tomographic measurement system at the Halden Boiling Water Reactor.

  • 14.
    Holcombe, Scott
    et al.
    Inst Energy Technol, OECD Halden Reactor Project, N-1751 Halden, Norway..
    Svärd, Staffan Jacobsson
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Hallstadius, Lars
    Westinghouse Elect Sweden AB, S-72163 Vasteras, Sweden..
    A Novel gamma emission tomography instrument for enhanced fuel characterization capabilities within the OECD Halden Reactor Project2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 837-845Article in journal (Refereed)
    Abstract [en]

    Gamma emission tomography is a method based on gamma-ray spectroscopy and tomographic reconstruction techniques, which can be used for rod-wise characterization of nuclear fuel assemblies without dismantling the fuel. By performing a large number of measurements of the gamma-ray flux intensity around a fuel assembly using a well-collimated gamma-ray detector, the internal source distribution in the assembly may be reconstructed using tomographic algorithms. If a spectroscopic detection system is used, different gamma-ray emitting isotopes can be selected for analysis, enabling nondestructive fuel characterization with respect to a variety of fuel parameters. In this paper, we describe a novel gamma emission tomography instrument, which has been designed, constructed and tested at the Halden Boiling Water Reactor (HBWR). The device will be used to characterize fuel assemblies irradiated in the HBWR as part of ongoing nuclear fuel research conducted within the OECD Halden Reactor Project (HRP). As compared to single-rod gamma scanning, where the fuel is dismantled and the gamma radiation from each rod is measured separately, handling time associated with characterizing the fuel can be significantly reduced when using the gamma emission tomography device. Furthermore, because gamma emission tomography enables rod-wise fuel characterization without dismantling, even instrumented experimental fuel assemblies may be characterized repeatedly throughout the fuel's lifetime, with limited risk of damaging the fuel or its instrumentation. Accordingly, the capabilities of fuel characterization within the OECD HRP are expected to be strongly enhanced by the deployment of this device. Here, the gamma-tomographic method and the experimental setup are demonstrated through experimental measurements of the fuel stack and gas plenum regions of a nine-rod HBWR fuel assembly configuration, where four rods had a burnup of approximately 26 MWd/kgUO(2) and five rods had a burnup of approximately 50 MWd/kgUO(2). Tomographic images are presented, which show the applicability for assessment of fission gas contents in the gas plena and of fission products in the fuel stack. Furthermore, neutron activation products are analyzed, which give additional information on construction material properties.

  • 15.
    Loberg, John
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Bejmer, Klaes-Håkan
    Blomgren, Jan
    Kierkegaard, Jesper
    Investigation of axial power gradients near a control rod tip2011In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 38, no 7, p. 1609-1615Article in journal (Refereed)
    Abstract [en]

    Control rod withdrawal in BWRs induces large power steps in the adjacent fuel assemblies. This paper investigates how well a 2D/3D method, e.g., CASMO5/SIMULATE5 computes axial pin power gradients adjacent to an asymmetrical control-rod tip in a BWR. The ability to predict pin power gradients accurately is important for safety considerations whereas large powers steps induced by control rod withdrawal can cause Pellet Cladding Interaction. The computation of axial pin power gradients axially around a control rod tip is a challenging task for any nodal code. On top of that, asymmetrical control rod handles are present in some BWR designs. The lattice code CASMO requires diagonal symmetry of all control rod parts. This introduces an error in computed pin power gradients that has been evaluated by Monte Carlo calculations. The results show that CASMO5/SIMULATE5, despite the asymmetrical control rod handle, is able to predict the axial pin power gradient within 1%/cm for axial nodal sizes of 15-3.68 cm. However, a nodal size of 3.68 cm still causes underestimations of pin power gradients compared with 1 cm nodes. Furthermore, if conventional node sizes are used, similar to 15 cm, pin power gradients can be underestimated by over 50% compared with 1 cm nodes. The detailed axial pin power profiles from MCNP are corroborated by measured gamma scan data on fuel rods irradiated adjacent to control rods.

  • 16.
    Loberg, John
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Bejmer, Klaes-Håkan
    Blomgren, Jan
    Kierkegaard, Jesper
    Investigation of Axial Power Gradients near an Unsymmetrical Control Blade TipIn: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
  • 17.
    Matsson, Ingvar
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics, Nuclear Physics.
    Grapengiesser, Björn
    The shut-down of the Barseback 1 BWR: A unique opportunity to measure the power distribution in nuclear fuel rods2006In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 33, no 13, p. 1091-1101Article in journal (Refereed)
    Abstract [en]

    Reactor poolside measurements of gamma radiation specific for the fission product La-140 (1596 keV) have been used for an experimental determination of axial power distributions in 55 nuclear fuel rods irradiated in the Barseback 1 BWR nuclear power plant. The measurements take advantage of the unique situation of a very short last reactor cycle of only three months due to the out-phasing of the reactor unit at November 30 1999. La-140 whose decay is controlled by the mother nuclide Ba-140 with the half-life 12.75 days reflects an average power distribution, representative for the latest weeks of core operation (in this case basically during November 1999). The measured intensities have been transformed into a 25 nodal representation to allow a precise and direct comparison with the corresponding calculated power distribution. The 55 rods were selected from two different fuel assemblies with average burn-ups of 1.9 and 9.7 MWd/ kgU, respectively (that is one fresh bundle and one slightly more than one cycle bundle). The stability and the linearity of the measurement system were evaluated. The linearity was checked using the two-source method. The stability was checked by recurrent measurements on a reference fuel rod. The results have been used in the validation of the pin power reconstruction model of Westinghouse 3D core simulator POLCA-7. The deviation between measured and calculated Ba-140 concentration (expressed as radial error) is typically a few percent on rod level. Results indicate that also Gd-rods are properly modelled over a broad range of conditions. It is indicated that predictions for fuel rods in their first month of operation are less accurate than for the rest of the rods.

  • 18.
    Noack, Klaus
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Moiseenko, Vladimir E
    Institute of Plasma Physics, National Science Center "Kharkiv Institute of Physics and Technology",Kharkiv, Ukraina.
    Ågren, Olov
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Hagnestål, Anders
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Neutronic model of a mirror based fusion-fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification2011In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 38, no 2-3, p. 578-589Article in journal (Refereed)
    Abstract [en]

    The Georgia Institute of Technology has developed several design concepts of tokamak based fusion-fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a "fusion transmutation of waste reactor". It was designed for a criticality keff <= 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GWth requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBRcycle = 1.9 and a first wall lifetime of 12-16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead-bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality keff <= 0.97 and produces less fission power of 1.5 GWth. Its main performance parameters turn out substantially better.

  • 19.
    Noack, Klaus
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Ågren, Olov
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Moiseenko, V. E.
    Hagnestål, Anders
    Uppsala University, Disciplinary Domain of Science and Technology, Technology, Department of Engineering Sciences, Electricity.
    Comments on the power amplification factor of a driven subcritical system2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 59, p. 261-266Article in journal (Refereed)
    Abstract [en]

    The power amplification factor PAF of a driven subcritical system is defined as the ratio of the fission power output of the blanket to the power which the driver must deliver to sustain its neutron source intensity. This parameter decisively determines the effectiveness of the whole system independent of its special purpose as energy amplifier or as transmutation facility. The present note derives a refined analytical expression for the PAF which reveals more physical details than the expressions given by other authors. Moreover, the traditionally used forms of the static reactor eigenvalue equation and of its adjoint equation are rewritten for subcritical systems and used in the derivation of the expression for the PAF. The derived formula and the modified eigenvalue equations are discussed.

  • 20.
    Qvist, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Optimization method for the design of hexagonal fuel assemblies2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 75, p. 498-506Article in journal (Refereed)
    Abstract [en]

    The duct wall and inter-assembly gap make up 5–15% of the volume of a typical fast reactor core and these components have a profound impact on the system neutron economy. In this paper, a methodology for the design of optimum hexagonal fuel assembly geometries was developed. For each nuclear reactor core made up of ducted assemblies there exists a unique optimum solution of duct wall thickness and inter-assembly gap, where these components have the minimum impact on the core neutron balance while adhering to applicable structural constraints. The assembly duct wall must maintain its structural integrity and intended function while being exposed to a harsh environment of pressure, temperature and neutron fluence causing elastic and inelastic deflections and swelling. Analytical expressions, appli- cable to any internally pressurized hexagonal structure, were defined for the peak stress and elastic wall deflection. Detailed analysis of fuel assembly duct designs requires finite element code analysis, radial bowing analysis codes and the full temperature, flux and stress distribution over the lifetime of the assembly in the core to accurately estimate creep deformation. The simple analytical methodology pre- sented in this paper can provide a good initial guess for an optimal geometry to be iteratively improved and refined using more advanced codes and methods.

  • 21.
    Qvist, Staffan A.
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Nuclear Physics. Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Hellesen, Carl
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Gradecka, Malwina
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Dubberley, Allen E.
    Gen Elect Adv Reactor Syst Dept, Sunnyvale, CA USA..
    Fanning, Thomas
    Argonne Natl Lab, Nucl Engn Div, Argonne, IL 60439 USA..
    Greenspan, Ehud
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Tailoring the response of Autonomous Reactivity Control (ARC) systems2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 99, p. 383-398Article in journal (Refereed)
    Abstract [en]

    The Autonomous Reactivity Control (ARC) system was developed to ensure inherent safety of Generation IV reactors while having a minimal impact on reactor performance and economic viability. In this study we present the transient response of fast reactor cores to postulated accident scenarios with and without ARC systems installed. Using a combination of analytical methods and numerical simulation, the principles of ARC system design that assure stability and avoids oscillatory behavior have been identified. A comprehensive transient analysis study for ARC-equipped cores, including a series of Unprotected Loss of Flow (ULOF) and Unprotected Loss of Heat Sink (ULOHS) simulations, were performed for Argonne National Laboratory (ANL) Advanced Burner Reactor (ABR) designs. With carefully designed ARC-systems installed in the fuel assemblies, the cores exhibit a smooth non-oscillatory transition to stabilization at acceptable temperatures following all postulated transients. To avoid oscillations in power and temperature, the reactivity introduced per degree of temperature change in the ARC system needs to be kept below a certain threshold the value of which is system dependent, the temperature span of actuation needs to be as large as possible.

  • 22.
    Qvist, Staffan
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Greenspan, Ehud
    The ADOPT code for automated fast reactor core design2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 71, p. 23-36Article in journal (Refereed)
    Abstract [en]

    The Assembly Design and OPTimization code (ADOPT) is a comprehensive computer code written to automate the process of designing and analyzing fast reactor fuel assemblies and cores. It finds a fuel assembly design that maximizes the fuel volume fraction in the core while adhering to set constraints for all component temperatures, pressure drop, coolant velocity and structural integrity limits, subjected to a specified assembly peak power level. ADOPT can be used very effectively as the first step-in the design process of fast reactor cores that offer the maximum possible breeding ratio, which is proportional to the fuel volume fraction. To design fast reactor cores with different objectives, one can start with a neutronic analysis to find material volume fractions that provide the sought core performance. ADOPT can then reverse-engineer a fuel assembly design with the desired volume fractions that abide by all the thermal-hydraulic and structural constraints. The code provides the necessary input files for a full core analysis to either SERPENT or MCNP neutron transport codes. Power and flux profiles from neutron transport calculations are then used to refine the ADOPT solution until a converged solution, considering thermal-hydraulics, structural mechanics and neutronics is achieved. 

  • 23.
    Qvist, Staffan
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Hou, Jason
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Greenspan, Ehud
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA..
    Design and performance of 2D and 3D-shuffled breed-and-burn cores2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 93-114Article in journal (Refereed)
    Abstract [en]

    The primary objective of this work is to find design approaches that will enable 3D fuel shuffling in stationary breed-and-burn (B&B) cores and to quantify the attainable reduction in peak DPA and change in additional performance characteristics going from conventional 2D to 3D fuel shuffling strategies. An additional objective is to establish the tradeoff between the minimum required DPA (displacements per atom) and average required bumup (fuel utilization) for B&B cores spanning a core power range from 1250 to 3500 MWth. It is found possible to design a B&B core fuelled with depleted uranium to have a peak radiation damage at or below 350 DPA when using 3D-shuffling. Relative to conventional 2D-shuffling, 3D-shuffling offers between 30% and 40% reduction in the peak DPA along with up to 30% increase in the average discharge bumup and, hence, in the depleted uranium utilization as well as significant increase in the core average and specific power density. Per DPA, the 3D shuffling option offers up to 60% higher uranium utilization. Even though 350 DPA is above the 200 DPA peak radiation damage HT9 steels were exposed to so far, it is below the 400 DPA advanced structural materials are expected to tolerate.

  • 24.
    Rochman, D.
    et al.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Leray, O.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Perret, G.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Vasiliev, A.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Ferroukhi, H.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Koning, Arjan J.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. IAEA, Nucl Data Sect, A-1400 Vienna, Austria..
    Re-evaluation of the thermal neutron capture cross section of Nd-1472016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 94, p. 612-617Article in journal (Refereed)
    Abstract [en]

    In this paper we are proposing a re-evaluation of the thermal-neutron induced capture cross section of Nd-147. A unique measurement exists from which this cross section was calculated in 1974. This original calculation is based on an assumed value for a specific gamma-ray fraction (called F-2), taken from the neighboring nucleus Nd-145. With the availability of reaction codes such as TALYS, such fraction can nowadays be calculated using specific reaction models and parameters. The new value of F-2 indicates a decrease of the thermal cross section by 45%, leading to 243 barns, instead of the 440 barns previously reported. This new cross section impacts the calculation of the number density for the well-known burn-up indicator Nd-148, but as shown, the change is close to the usual experimental uncertainties for the 148Nd number densities, thus having a limited impact on burn-up calculation.

  • 25.
    Rochman, D.
    et al.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Leray, O.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Vasiliev, A.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Ferroukhi, H.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Koning, Arjan J.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. IAEA, Nucl Data Sect, Vienna, Austria..
    Fleming, M.
    Culham Ctr Fus Energy, Abingdon, Oxon, England..
    Sublet, J. C.
    Culham Ctr Fus Energy, Abingdon, Oxon, England..
    A Bayesian Monte Carlo method for fission yield covariance information2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 95, p. 125-134Article in journal (Refereed)
    Abstract [en]

    The present work proposes a Bayesian method to combine theoretical fission yields with a set of reference data. These two sources of information are merged using a Monte Carlo process, and leads to a so-called Bayesian Monte Carlo update. Examples are presented for the independent fission yields of four major actinides, using the GEF code as a source of theoretical calculations and an evaluated library of fission yields for the reference data. The impact of the updated fission yields and their covariances is shown for two distinct applications: a UO2 pincell with burn-up up to 40 GWD/tHM and decay heat calculations of a thermal neutron pulse on U-235 and Pu-239.

  • 26.
    Rochman, D.
    et al.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Vasiliev, A.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Ferroukhi, H.
    Paul Scherrer Inst, Reactor Phys & Syst Behav Lab, Villigen, Switzerland..
    Zhu, T.
    Univ Florida, Gainesville, FL USA..
    van der Marck, S. C.
    Nucl Res & Consultancy Grp NRG, Petten, Netherlands..
    Koning, Arjan J.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics. IAEA, Nucl Data Sect, A-1400 Vienna, Austria.
    Nuclear data uncertainty for criticality-safety: Monte Carlo vs. linear perturbation2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 92, p. 150-160Article in journal (Refereed)
    Abstract [en]

    This work is presenting a comparison of results for different methods of uncertainty propagation due to nuclear data for 330 criticality-safety benchmarks. Covariance information is propagated to key using either Monte Carlo methods (NUSS: based on existing nuclear data covariances, and TMC: based on reaction model parameters) or sensitivity calculations from MCNP6 coupled with nuclear data covariances. We are showing that all three methods are globally equivalent for criticality calculations considering the two first moments of a distribution (average and standard deviation), but the Monte Carlo methods lead to actual probability distributions, where the third moment (skewness) should not be ignored for safety assessments.

  • 27. Suvdantsetseg, Erdenechimeg
    et al.
    Qvist, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Ehud, Greenspan
    Preliminary transient analysis of the Autonomous Reactivity Control system for fast reactors2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 77, p. 47-64Article in journal (Refereed)
    Abstract [en]

    A preliminary parametric dynamic response study of the Autonomous Reactivity Control (ARC) system is performed for a large fast reactor core subjected to the following postulated design-extension accident scenarios: unprotected overpower (UTOP), unprotected loss of flow (ULOF) and unprotected loss of heat sink (ULOHS). The results show that the ARC system could prevent fuel and cladding failure as well as coolant boiling during UTOP and can lower transient temperatures during ULOHS. However, in case of ULOF, the design space examined the ARC system did not prevent exceeding permissible cladding temperature and introduced instability. A more thorough analysis is recommended.

  • 28. Trost, Nico
    et al.
    Jiménez, Javier
    Lukarski, Dimitar
    Uppsala University, Disciplinary Domain of Science and Technology, Mathematics and Computer Science, Department of Information Technology, Division of Scientific Computing. Uppsala University, Disciplinary Domain of Science and Technology, Mathematics and Computer Science, Department of Information Technology, Computational Science.
    Sanchez, Victor
    Accelerating COBAYA3 on multi-core CPU and GPU systems using PARALUTION2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 82, p. 252-259Article in journal (Refereed)
  • 29.
    Verma, Vasudha
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Filliatre, Philippe
    CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache.
    Hellesen, Carl
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jammes, Christian
    CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache.
    Neutron flux monitoring with in-vessel fission chambers to detect an inadvertent control rod withdrawal in a sodium-cooled fast reactor2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 94, p. 487-493Article in journal (Refereed)
    Abstract [en]

    The neutron flux monitoring system forms an integral part of the safety design of a Generation IV sodium-cooled fast reactor. During the initial design phase of the neutron flux monitoring system, one needs to explore various detector locations and configurations. Diverse possibilities of the detector system installation should be studied for different locations in the reactor vessel in order to detect any perturbations in the core. In this paper, we investigate the possibility of placing fission chambers beyond the lateral neutron shield, ex-core but in-vessel and study the detectability of an inadvertent control rod withdrawal with these fission chambers. A generic core design of a Generation IV 1500 MWth French sodium-cooled fast reactor is used for the study, and calculations are performed with the Monte Carlo code SERPENT2. We propose certain design changes that are needed to be incorporated, w.r.t. the facilitation of neutron transport to this ex-core location.

    We are able to show that there is a detectable signature in the fission chambers following an inadvertent control rod withdrawal in the core. The equally-spaced azimuthal detectors are able to follow changes in the neutron flux distribution in the core. This study helps us to analyze multiple detector locations and give the general trends for monitoring indications to detect any perturbations in the core.

  • 30.
    Willman, Christofer
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics.
    Håkansson, Ane
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics. Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Osifo, Otasowie
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics.
    Bäcklin, Anders
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics. Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Nuclear and Particle Physics. Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes2006In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 33, no 9, p. 766-773Article in journal (Refereed)
    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.

  • 31.
    Wolniewicz, Peter
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Hellesen, Carl
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jansson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Håkansson, Ane
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Österlund, Michael
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Detecting neutron spectrum perturbations due to coolant density changes in a small lead-cooled fast nuclear reactor2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 58, p. 102-109Article in journal (Refereed)
    Abstract [en]

    The lead-cooled fast reactor (LFR) is one of the nuclear reactor technologies proposed by the Generation IV International Forum (GIF). The lead coolant allows for inherent safety properties attractive from a nuclear safety point of view, but issues related to corrosion of structural materials and the possible positive coolant reactivity coefficient must be addressed before LFRs can be commercially viable. As an example, a small crack in e.g. a heat exchanger can generate a more or less homogeneous distribution of bubbles in the coolant (void) which if unnoticed, has the potential to cause criticality issues. This fact motivated an investigation of a methodology to detect such voids.

    The suggested methodology is based on measurements of the “slow” and “fast” parts of the neutron spectrum because these parts respond in different ways to voiding. For detection, it is tentatively assumed that fission chambers loaded with U-235 and Pu-239, respectively, are deployed. To investigate the methodology according to sensitivity and precision, a number of scenarios have been simulated and analysed using the core simulator Serpent.

    The results show that the methodology yields a sensitivity of 3% for each per cent unit of void. Assuming typical detection limits of a few per cent this implies the possibility to detect voids down to the order of 1%. From these studies it was also concluded that the positioning of the detectors relative the reactor core is crucial, which may be useful input during the design phase of a reactor in order to achieve an efficient monitoring system.

  • 32.
    Wolniewicz, Peter
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Håkansson, Ane
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jansson, Peter
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Detection of coolant void in lead-cooled fast reactors2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 1096-1103Article in journal (Refereed)
    Abstract [en]

    Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242Pu and 235U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor.

    In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology.

  • 33. Åberg Lindell, M.
    et al.
    Grape, S.
    Håkansson, A.
    Svärd, S. Jacobsson
    Erratum to “Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology” [Ann. Nucl. Energy 62 (2013) 390–397]2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 66, no Supplement C, p. 61-62Article in journal (Refereed)
  • 34.
    Åberg Lindell, Matilda
    et al.
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Grape, Sophie
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Håkansson, Ane
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Jacobsson Svärd, Staffan
    Uppsala University, Disciplinary Domain of Science and Technology, Physics, Department of Physics and Astronomy, Applied Nuclear Physics.
    Assessment of proliferation resistances of aqueous reprocessing techniques using the TOPS methodology2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 62, p. 390-397Article in journal (Refereed)
    Abstract [en]

    The aim of this study is to assess and compare the proliferation resistances (PR) of three possible Generation IV lead-cooled fast reactor fuel cycles, involving the reprocessing techniques Purex, Ganex and a combination of Purex, Diamex and Sanex, respectively. The examined fuel cycle stages are reactor operation, reprocessing and fuel fabrication. The TOPS methodology has been chosen for the PR assessment, and the only threat studied is the case where a technically advanced state diverts nuclear material covertly.

    According to the TOPS methodology, the facilities have been divided into segments, here roughly representing the different forms of nuclear material occurring in each examined fuel cycle stage. For each segment, various proliferation barriers have been assessed.

    The results make it possible to pinpoint where the facilities can be improved. The results show that the proliferation resistance of a fuel cycle involving recycling of minor actinides is higher than for the traditional Purex reprocessing cycle. Furthermore, for the purpose of nuclear safeguards, group actinide extraction should be preferred over reprocessing options where pure plutonium streams occur. This is due to the fact that a solution containing minor actinides is less attractive to a proliferator than a pure Pu solution. Thus, the safeguards analysis speaks in favor of Ganex as opposed to the Purex process.

1 - 34 of 34
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