| 31. |
- Alhassan, Erwin, et al.
(författare)
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On the use of integral experiments for uncertainty reduction of reactor macroscopic parameters within the TMC methodology
- 2016
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Ingår i: Progress in nuclear energy (New series). - : Elsevier BV. - 0149-1970 .- 1878-4224. ; 88, s. 43-52
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Tidskriftsartikel (refereegranskat)abstract
- The current nuclear data uncertainties observed in reactor safety parameters for some nuclides call for safety concerns especially with respect to the design of GEN-IV reactors and must therefore be reduced significantly. In this work, uncertainty reduction using criticality benchmark experiments within the Total Monte Carlo methodology is presented. Random nuclear data libraries generated are processed and used to analyze a set of criticality benchmarks. Since the calculated results for each random nuclear data used are different, an algorithm was used to select (or assign weights to) the libraries which give a good description of experimental data for the analyses of the benchmarks. The selected or weighted libraries were then used to analyze the ELECTRA reactor. By using random nuclear data libraries constrained with only differential experimental data as our prior, the uncertainties observed were further reduced by constraining the files with integral experimental data to obtain a posteriori uncertainties on the k(eff). Two approaches are presented and compared: a binary accept/reject and a method of assigning file weights based on the likelihood function. Significant reductions in (PU)-P-239 and Pb-208 nuclear data uncertainties in the k(eff) were observed after implementing the two methods with some criticality benchmarks for the ELELIRA reactor. (C) 2015 Elsevier Ltd. All rights reserved.
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| 32. |
- Alhassan, Erwin, et al.
(författare)
-
Reducing A Priori 239Pu Nuclear Data Uncertainty In The Keff Using A Set Of Criticality Benchmarks With Different Nuclear Data Libraries
- 2015
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Konferensbidrag (övrigt vetenskapligt/konstnärligt)abstract
- In the Total Monte Carlo (TMC) method [1] developed at the Nuclear Research and Consultancy Group for nuclear data uncertainty propagation, model calculations are compared with differential experimental data and a specific a priori uncertainty is assigned to each model parameter. By varying the model parameters all together within model parameter uncertainties, a full covariance matrix is obtained with its off diagonal elements if desired [1]. In this way, differential experimental data serve as a constraint for the model parameters used in the TALYS nuclear reactions code for the production of random nuclear data files. These files are processed into usable formats and used in transport codes for reactor calculations and for uncertainty propagation to reactor macroscopic parameters of interest. Even though differential experimental data together with their uncertainties are included (implicitly) in the production of these random nuclear data files in the TMC method, wide spreads in parameter distributions have been observed, leading to large uncertainties in reactor parameters for some nuclides for the European Lead cooled Training Reactor [2]. Due to safety concerns and the development of GEN-IV reactors with their challenging technological goals, the present uncertainties should be reduced significantly if the benefits from advances in modelling and simulations are to be utilized fully [3]. In Ref.[4], a binary accept/reject approach and a more rigorous method of assigning file weights based on the likelihood function were proposed and presented for reducing nuclear data uncertainties using a set of integral benchmarks obtained from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP). These methods are depended on the reference nuclear data library used, the combined benchmark uncertainty and the relevance of each benchmark for reducing nuclear data uncertainties for a particular reactor system. Since each nuclear data library normally comes with its own nominal values and covariance matrices, reactor calculations and uncertainties computed with these libraries differ from library to library. In this work, we apply the binary accept/reject approach and the method of assigning file weights based on the likelihood function for reducing a priori 239Pu nuclear data uncertainties for the European Lead Cooled Training Reactor (ELECTRA) using a set of criticality benchmarks. Prior and posterior uncertainties computed for ELECTRA using ENDF/B-VII.1, JEFF-3.2 and JENDL-4.0 are compared after including experimental information from over 10 benchmarks.[1] A.J. Koning and D. Rochman, Modern Nuclear Data Evaluation with the TALYS Code System. Nuclear Data Sheets 113 (2012) 2841-2934. [2] E. Alhassan, H. Sjöstrand, P. Helgesson, A. J. Koning, M. Österlund, S. Pomp, D. Rochman, Uncertainty and correlation analysis of lead nuclear data on reactor parameters for the European Lead Cooled Training reactor (ELECTRA). Annals of Nuclear Energy 75 (2015) 26-37. [3] G. Palmiotti, M. Salvatores, G. Aliberti, H. Hiruta, R. McKnight, P. Oblozinsky, W. Yang, A global approach to the physics validation of simulation codes for future nuclear systems, Annals of Nuclear Energy 36 (3) (2009) 355-361. [4] E. Alhassan, H. Sjöstrand, J. Duan, P. Helgesson, S. Pomp, M. Österlund, D. Rochman, A.J. Koning, Selecting benchmarks for reactor calculations: In proc. PHYSOR 2014 - The Role of Reactor Physics toward a Sustainable Future, kyoto, Japan, Sep. 28 - 3 Oct. (2014).
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| 33. |
- Alhassan, Erwin, et al.
(författare)
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Selecting benchmarks for reactor calculations
- 2014
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Ingår i: PHYSOR 2014 - The Role of Reactor Physics toward a Sustainable Future.
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Konferensbidrag (refereegranskat)abstract
- Criticality, reactor physics, fusion and shielding benchmarks are expected to play important roles in GENIV design, safety analysis and in the validation of analytical tools used to design these reactors. For existing reactor technology, benchmarks are used to validate computer codes and test nuclear data libraries. However the selection of these benchmarks are usually done by visual inspection which is dependent on the expertise and the experience of the user and there by resulting in a user bias in the process. In this paper we present a method for the selection of these benchmarks for reactor applications based on Total Monte Carlo (TMC). Similarities betweenan application case and one or several benchmarks are quantified using the correlation coefficient. Based on the method, we also propose an approach for reducing nuclear data uncertainty using integral benchmark experiments as an additional constrain on nuclear reaction models: a binary accept/reject criterion. Finally, the method was applied to a full Lead Fast Reactor core and a set of criticality benchmarks.
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| 34. |
- Alhassan, Erwin, et al.
(författare)
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Selecting benchmarks for reactor simulations : an application to a Lead Fast Reactor
- 2016
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 158-169
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Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 35. |
- Alhassan, Erwin, et al.
(författare)
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Uncertainty analysis of Lead cross sections on reactor safety for ELECTRA
- 2016
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Ingår i: SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo. - Les Ulis, France : EDP Sciences.
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Konferensbidrag (refereegranskat)abstract
- The Total Monte Carlo (TMC) method was used in this study to assess the impact of Pb-206, 207 and 208 nucleardata uncertainties on k-eff , beta-eff, coolant temperature coefficient, the coolant void worth for the ELECTRA reactor. Relatively large uncertainties were observed in the k-eff and the coolant void worth for all the isotopes with significant contribution coming from Pb-208 nuclear data. The large Pb-208 nuclear data uncertainty observed was further investigated by studying the impact of partial channels on the k-eff and beta-eff. Various sections of ENDF file: elasticscattering (n,el), inelastic scattering (n,inl), neutron capture (n,gamma), (n,2n), resonance parameters and the angular distribution were varied randomly and distributions in k-eff and beta-eff obtained. The dominant contributions to the uncertainty in the k-eff from Pb-208 came from uncertainties in the resonance parameters; however, elastic scattering cross section and the angular distribution also had significant impact. The impact of nuclear data uncertainties on the beta-eff was observed to be small.
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| 36. |
- Alhassan, Erwin, et al.
(författare)
-
Uncertainty and correlation analysis of lead nuclear data on reactor parameters for the European Lead Cooled Training Reactor
- 2015
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 26-37
-
Tidskriftsartikel (refereegranskat)abstract
- 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.
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| 37. |
- Andersson, Pernilla, et al.
(författare)
-
An upgrade of the SCANDAL facility for neutron scattering measurements at 175 MeV
- 2009
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Rapport (övrigt vetenskapligt/konstnärligt)abstract
- The experimental setup SCANDAL (SCAttered Nucleon Detection AssembLy) at the The Svedberg Laboratory (TSL), previously used for measurements of the differential cross section of elastic and inelastic neutron scattering in the 50 – 130 MeV range, has recently been upgraded with new Na doped CsI scintillating detectors for measurements at 175 MeV. The performance of the new setup is described and illustrated by the early steps in the analysis of the first experimental campaign, carried out in January and February 2009.
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| 38. |
- Andersson, P., et al.
(författare)
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An upgrade of the SCANDAL setup for measurements of elastic neutron scattering at 175 MeV
- 2010
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Ingår i: RADIATION MEASUREMENTS. - : Elsevier BV. ; 45:10, s. 1142-1144
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Konferensbidrag (refereegranskat)abstract
- The experimental setup SCANDAL used for measurements of the differential cross section for elastic and inelastic neutron scattering, has recently been upgraded with larger CsI scintillating detectors to enable measurements at energies up to 175 MeV. Measurements on Fe. Bi and Si have been carried out using the quasi mono-energetic neutron beam at the The Svedberg Laboratory, and data is under analysis. The experimental setup can be used for measurements on a wide range of target nuclei, including C and O, which are important for dosimetry applications. SCANDAL can also run in proton mode, for measurements of the (n,p) reaction. This paper describes the new experimental setup, and reports on its properties, such as energy resolution.
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| 39. |
- Andersson, Peter, 1981-
(författare)
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Fast-Neutron Tomography using a Mobile Neutron Generator for Assessment of Steam-Water Distributions in Two-Phase Flows
- 2014
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- This thesis describes the measurement technique of fast-neutron tomography for assessing spatial distributions of steam and water in two-phase flows. This so-called void distribution is of importance both for safe operation and for efficient use of the fuel in light water reactors, which compose the majority of the world’s commercial nuclear reactors. The technique is aimed for usage at thermal-hydraulic test loops, where heated two-phase flows are being investigated under reactor-relevant conditions.By deploying portable neutron generators in transmission tomography, the technique becomes applicable to stationary objects, such as thermal-hydraulic test loops. Fast neutrons have the advantage of high transmission through metallic structures while simultaneously being relatively sensitive to the water/void content. However, there are also challenges, such as the relatively low yield of commercially available fast-neutron generators, the tendency of fast neutrons to scatter in the interactions with materials and the relatively low efficiency encountered in fast-neutron detection.The thesis describes the design of a prototype instrument, FANTOM, which has been assembled and demonstrated. The main design parameters have been optimized to achieve maximal signal count rate in the detector elements, while simultaneously reaching an image unsharpness of ≤0.5 mm. Radiographic projections recorded with the assembled instrument are presented, and the performance parameters of FANTOM are deduced.Furthermore, tomographic reconstruction methods for axially symmetric objects, which is relevant for some test loops, have been developed and demonstrated on measured data from three test objects. The attenuation distribution was reconstructed with a radial resolution of 0.5 mm and an RMS error of 0.02 cm-1, based on data recorded using an effective measurement time of 3.5 hours per object. For a thermal-hydraulic test loop, this can give a useful indication of the flow mode, but further development is desired to improve the precision of the measurements.Instrument upgrades are foreseen by introducing a more powerful neutron generator and by adding detector elements, speeding up the data collection by several orders of magnitude and allowing for higher precision data. The requirements and performance of an instrument for assessment of arbitrary non-symmetric test loops is discussed, based on simulations.
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| 40. |
- Andersson, Peter, 1981-
(författare)
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Optimization of Equipment for Tomographic Measurements of Void Distributions using Fast Neutrons
- 2011
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Licentiatavhandling (övrigt vetenskapligt/konstnärligt)abstract
- This licentiate thesis describes a novel nondestructive measuring technique for determiningspatial distributions of two-phase water flows. In Boiling Water Reactors, which compose themajority of the world's commercial nuclear reactors, this so called void distribution is of importance for safe operation.The presented measurement technique relies on fast neutron transmission tomography using portable neutron generators. Varying hardware options for such an instrument based on this technique and a prototype instrument, which is under construction, are described. The main design parameters are detailed and motivated from a performance point of view. A Paretomultiple objective optimization of the count rate and image unsharpness is presented. The resulting instrument design comprises an array of plastic scintillators for neutron detection. Such detector elements allow for spectroscopic data acquisition and subsequent reduction of background events at low energy by means of introducing an energy threshold in the analysis.The thesis includes two papers: In paper I, the recoil proton energy deposition distribution resulting from the interaction of the incoming neutrons is investigated for thin plastic scintillator elements. It is shown that the recoil proton losses have a large effect on the pulse height distribution and the intrinsic neutron detection efficiency is calculated for varying energy thresholds.In paper II the performance of the planned FANTOM device is investigated using the particle transport code MCNP5. An axially symmetric phantom void distribution is modeled and there construction is compared with the correct solution. According to the solutions, the phantom model can be reconstructed with 10 equal size ring-shaped picture elements, with a precision of better than 5 void percent units using a deuterium-tritium neutron generator with a yield of 3 · 107 neutrons per second and a measurement time of 13 h. However, it should be noted that commercial neutron generators with a factor of 103 higher yields exist and that the measurement time could decrease to less than a minute if such a neutron generator would beutilized.
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