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Sökning: L773:0306 4549 OR L773:1873 2100 > (2015-2019)

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1.
  • Alhassan, Erwin, et al. (författare)
  • Benchmark selection methodology for reactor calculations and nuclear data uncertainty reduction
  • 2015
  • Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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2.
  • Alhassan, Erwin, et al. (författare)
  • Selecting benchmarks for reactor simulations : an application to a Lead Fast Reactor
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 158-169
  • 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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3.
  • 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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4.
  • Andersson, Peter, 1981-, et al. (författare)
  • A computerized method (UPPREC) for quantitative analysis of irradiated nuclear fuel assemblies with gamma emission tomography at the Halden reactor
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 88-97
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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5.
  • Basso, Simone, et al. (författare)
  • Effectiveness of the debris bed self-leveling under severe accident conditions
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 95, s. 75-85
  • Tidskriftsartikel (refereegranskat)abstract
    • Melt fragmentation, quenching and long term coolability in a deep pool of water under the reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. The success of such strategy is contingent upon the natural circulation effectiveness in removing the decay heat generated in the porous debris bed. The maximum height of the bed is one of the important factors which affect the debris coolability. The two-phase flow within the bed generates mechanical energy which can change the geometry of the debris bed by the "self-leveling" phenomenon. In this work.we developed an approach to modeling of the self-leveling phenomenon. Sensitivity analysis was carried out to rank the importance of the model uncertainties and uncertain input parameters i.e. the conditions of the accident scenario and the debris bed properties. The results provided some useful insights for further improvement of the model and reduction of the output uncertainties through separate-effect experimental studies. Finally, we assessed the self-leveling effectiveness, quantified its uncertainties in prototypic severe accident conditions and demonstrated that the effect of self-leveling phenomenon is robust with respect to the considered input uncertainties.
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6.
  • Bechta, Sevostian, et al. (författare)
  • On the EU-Japan roadmap for experimental research on corium behavior
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 124, s. 541-547
  • Tidskriftsartikel (refereegranskat)abstract
    • A joint research roadmap between Europe and Japan has been developed in severe accident field of light water reactors, focusing particularly on reactor core melt (corium) behavior. The development of this roadmap is one of the main targets of the ongoing EU project SAFEST. This paper presents information about ongoing severe accident studies in the area of corium behavior, rationales and comparison of research priorities identified in different projects and documents, expert ranking of safety issues, and finally the research areas and topics and their priorities suggested for the EU Japan roadmap and future bilateral collaborations. These results provide useful guidelines for (i) assessment of long-term goals and proposals for experimental support needed for proper understanding, interpretation and learning lessons of the Fukushima accident; (ii) analysis of severe accident phenomena; (iii) development of accident prevention and mitigation strategies, and corresponding technical measures; (iv) study of corium samples in European and Japanese laboratories; and (v) preparation of Fukushima site decommissioning.
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7.
  • Bortot, Sara, et al. (författare)
  • BELLA : a multi-point dynamics code for safety-informed design of fast reactors
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 85, s. 228-235
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS-1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment. (C) 2015 Elsevier Ltd. All rights reserved.
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8.
  • Chernikova, Dina, 1982, et al. (författare)
  • Novel passive and active tungsten-based identifiers for maintaining the continuity of knowledge of spent nuclear fuel copper canisters
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 219-227
  • Tidskriftsartikel (refereegranskat)abstract
    • A new approach to provide a long-term safeguards identification of spent nuclear fuel containers, in particular copper canisters, is presented in this paper. The approach proposes the use of a tungsten insert marked with a binary code and placed inside the container. The insert is read with a combination of two independent techniques, radiation and ultrasonic measurements, in order to get a unique identification of the cask. Passive and active versions of the tag are considered. The passive version makes use of the radiation coming from the spent nuclear fuel itself. The active version of the tag is based on the use of an artificially introduced mixture of α-emitting isotopes, such as 241Am with materials, 11B and 23Na, which easily undergo α-induced reactions with emission of specific γ-lines, 2313 keV and 1809 keV, respectively. The paper discusses results of the radiation and ultrasonic measurements and Monte-Carlo evaluations as the first proof of the concept. The results of the investigations show the strong potential for this concept to maintain the continuity of knowledge of spent nuclear fuel copper canisters for a time scale up to a few thousands years without compromising the environmental safety of the casks.
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9.
  • Davour, Anna, et al. (författare)
  • Applying image analysis techniques to tomographic images of irradiated nuclear fuel assemblies
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 96, s. 223-229
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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10.
  • Demazière, C., et al. (författare)
  • Development of three-dimensional capabilities for modelling stationary fluctuations in nuclear reactor cores
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 84, s. 19-30
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper presents the development of a numerical tool meant at modelling the effect of stationary fluctuations in nuclear cores for systems cooled with either liquid water or boiling water. The originating fluctuations are defined for the variables describing the boundary conditions of the system, i.e. inlet velocity, inlet enthalpy, and outlet pressure. The tool then determines in the frequency domain the three-dimensional distributions within the core of the corresponding fluctuations in neutron flux, coolant density, coolant velocity, coolant enthalpy, and fuel temperature. The tool is thus based on the simultaneous modelling of neutron transport, fluid dynamics, and heat transfer in a truly integrated and fully coupled manner. The modelling of neutron transport relies on the two-group diffusion approximation and a spatial discretization based on finite differences. The modelling of fluid dynamics is performed using the Homogeneous Equilibrium Model, with a void fraction correction based on a pre-computed distribution of the static slip ratio (when two-phase flow conditions are encountered). Heat conduction in the fuel pins is also accounted for, and the heat transfer between the fuel pins and the coolant is modelled also using a pre-computed distribution of the heat transfer coefficient. The spatial discretization of the fluid dynamic and heat transfer problems is carried out using finite volumes. The tool, currently entirely Matlab based, requires minimal input data, mostly in form of the three-dimensional distributions of the macroscopic cross-sections and their relative dependence on coolant density and fuel temperature, the point-kinetic parameters of the core, as well as the three-dimensional distribution of the slip ratio (in case of two-phase flow conditions) and of the heat transfer coefficient. Such data can be provided by any static core simulator that thus needs to be run prior to using the present tool. In addition to briefly summarizing the different test cases used to verify the code, the paper also presents the results of simulations performed for a typical Pressurized Water Reactor and for a typical Boiling Water Reactor, as illustrations of the capabilities of the tool. 
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11.
  • Dokhane, A., et al. (författare)
  • Analysis of Oskarshamn-2 stability event using TRACE/SIMULATE-3K and comparison to TRACE/PARCS and SIMULATE-3K stand-alone
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 102, s. 190-199
  • Tidskriftsartikel (refereegranskat)abstract
    • With the goal to enhance the capability to perform best-estimate simulations of Light Water Reactors (LWRs) transients, with strong coupling between core neutronics and plant thermal-hydraulic, a coupling between TRACE and SIMULATE-3K (TS3K) was developed in collaboration between PSI and Studsvik for analyses involving interactions between system and core. In order to verify the coupling scheme and the coupled code capabilities to simulate complex transients, the OECD/NEA Oskarshmn-2 (O-2) Stability benchmark was modeled with the coupled code TS3K. The main goal of this paper is to present TS3K analyses of the Oskarshamn-2 stability event, noting that this constitutes the first reported assessment of this code system for a BWR stability problem. A systematic analysis is carried out using different time-space discretization schemes in order to identify an optimized methodology to simulate correctly the O-2 stability event. In this context, the TS3K results are compared to the available benchmark data both for steady-state and transient conditions. The results show that using a refined model in space and time, the TS3K model can successfully capture the entire behavior of the transient qualitatively, i.e. onset of the instability with growing oscillation amplitudes, as well as quantitatively, i.e. Decay Ratio and resonance frequency. In addition, the results are compared also to those obtained using TRACE/PARCS and S3K stand-alone, which allows a systematic comparison between different codes.
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12.
  • Dufek, Jan, et al. (författare)
  • Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 94, s. 415-421
  • Tidskriftsartikel (refereegranskat)abstract
    • Previous studies raised a question about the level of a possible correlation of errors in the cumulative Monte Carlo fission source and the fundamental-mode eigenvector of the fission matrix. A number of new methods tally the fission matrix during the actual Monte Carlo criticality calculation, and use its fundamental-mode eigenvector for various tasks. The methods assume the fission matrix eigenvector is a better representation of the fission source distribution than the actual Monte Carlo fission source, although the fission matrix and its eigenvectors do contain statistical and other errors. A recent study showed that the eigenvector could be used for an unbiased estimation of errors in the cumulative fission source if the errors in the eigenvector and the cumulative fission source were not correlated. Here we present new numerical study results that answer the question about the level of the possible error correlation. The results may be of importance to all methods that use the fission matrix. New numerical tests show that the error correlation is present at a level which strongly depends on properties of the spatial mesh used for tallying the fission matrix. The error correlation is relatively strong when the mesh is coarse, while the correlation weakens as the mesh gets finer. We suggest that the coarseness of the mesh is measured in terms of the value of the largest element in the tallied fission matrix as that way accounts for the mesh as well as system properties. In our test simulations, we observe only negligible error correlations when the value of the largest element in the fission matrix is about 0.1. Relatively strong error correlations appear when the value of the largest element in the fission matrix raises above about 0.5. We also study the effect of the error correlations on accuracy of the eigenvector-based error estimator. The numerical tests show that the eigenvector-based estimator consistently underestimates the errors in the cumulative fission source when a strong correlation is present between the errors in the fission matrix eigenvector and the cumulative fission source (i.e., when the mesh is too coarse). The error estimates are distributed around the real error value when the mesh is sufficiently fine.
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13.
  • Dufek, Jan, 1978-, et al. (författare)
  • Monte Carlo criticality calculations accelerated by a growing neutron population
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 94, s. 16-21
  • Tidskriftsartikel (refereegranskat)abstract
    • We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. As the efficiency of Monte Carlo criticality simulations is sensitive to the selected neutron population size, the method attempts to achieve the acceleration via on-the-fly control of the neutron population size. The neutron population size is gradually increased over successive criticality cycles so that the fission source bias amounts to a specific fraction of the total error in the cumulative fission source. An optimal setting then gives a reasonably small neutron population size, allowing for an efficient source iteration; at the same time the neutron population size is chosen large enough to ensure a sufficiently small source bias, such that does not limit accuracy of the simulation.
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14.
  • Dykin, Victor, 1985, et al. (författare)
  • Predictive BWR core stability using feedback reactivity coefficients projected on neutronic eigenmodes
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 124, s. 1-8
  • Tidskriftsartikel (refereegranskat)abstract
    • The determination of the stability properties of Boiling Water Reactors usually rely on performing many time-dependent calculations for various combinations of values for the core power and the core flow. The aim of such calculations is to estimate the variation of the Decay Ratio in the core power/flow operating map, from which possible exclusion areas are defined. This paper demonstrates using a Reduced Order Model that the stability properties of a core with respect to global and regional oscillations are entirely determined by the projection of the feedback reactivity coefficients onto pairs of neutronic eigenmodes and their adjoint functions. This means that such projections inherently contain all information about the stability properties and their examination is sufficient to characterize the stability of a core. Most notably, the relative contributions of each fuel assembly to the core-wise projections give an indication to the core designer about the fuel assemblies possibly destabilizing the core. The core designer could thereafter improve core stability by either moving such assemblies to other locations or use another fuel assembly design. Although the method could be used independently of detailed stability calculations, the approach detailed in this study provides a more qualitative than quantitative core stability evaluation. This means that the method is most efficient if the stability features of a reference core are known.
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15.
  • Dykin, Victor, 1985, et al. (författare)
  • Remark on the neutron noise induced by propagating perturbations in an MSR
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 90, s. 93-105
  • Tidskriftsartikel (refereegranskat)abstract
    • The neutron noise induced by propagating perturbations in a simple model of a Molten Salt Reactor (MSR) is calculated and analyzed using one/two-group diffusion theory. The novelty, as compared to previous works, is that the noise source includes also the fluctuations of the fission cross sections and the fluid velocity, in addition to the previous case when only the fluctuations of the absorption cross section were accounted for. Another novelty is that the solution is obtained through the matrix Green's function of the flux and precursor equations, these two being kept separate. Inclusion of each of these two new noise sources leads to a structure of the noise source, and hence also that of the neutron noise, which is conceptually different from the case when only the fluctuations of the absorption cross sections are treated, with some surprising features. The use of the matrix Green's function is advantageous to understand the new features, and it helps to point out some new aspects of the neutron noise even in traditional systems, which have not been noticed before. The results contribute to the understanding and interpretation of the neutron noise in MSRs. (C) 2015 Elsevier Ltd. All rights reserved.
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16.
  • Fichot, F., et al. (författare)
  • Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 119, s. 36-45
  • Tidskriftsartikel (refereegranskat)abstract
    • The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactor below 500 MWe (like VVER440)) and is part of the SAMG strategies for some Gen III + PWRs of higher power like the AP1000 or the APR1400. However, for high power reactors, estimations using current level of conservatism show that RPV failure caused by thermo-mechanical rupture takes place in some cases. A better estimation of the residual risk (probability of cases with vessel rupture) requires the use of models with a lower level of conservatism. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness of the demonstration was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 MW/m2) whereas the first type provides the lowest heat fluxes (around 500 MW/m2) but is not realistic due to the non-miscibility of steel with UO2. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes in many cases are well above 1 MW/m2 which could reduce the residual thickness of the vessel considerably and threaten its integrity. Therefore, it is clear that the safety demonstration of IVR for high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. It also requires an accurate mechanical analysis of the ablated vessel. The current approach followed by most experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena (such as transient effects) and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Currently, the acceptance criteria of a safety demonstration for IVR may be differently defined from one country to the other and the differences should be further discussed to reach harmonization on this important topic. This includes the accident scenarios to be considered in the demonstration and the modelling of the phenomena in the vessel. Such harmonization is one of the goals of IVMR project. A revised methodology is proposed, where the safety criterion is based not only on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in current approaches but also on the minimum vessel thickness reached after ablation and the maximum integral loads that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion may be used in both probabilistic and deterministic approaches, whereas the current approaches are mostly deterministic (with deterministic calculations used only for estimates of uncertainty ranges of input parameters).
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17.
  • Gallego Marcos, Ignacio, et al. (författare)
  • Modelling of pool stratification and mixing induced by steam injectionthrough blowdown pipes
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 112, s. 624-639
  • Tidskriftsartikel (refereegranskat)abstract
    • Containment overpressure is prevented in a Boiling Water Reactor (BWR) by condensing steam into thepressure suppression pool. Steam condensation is a source of heat and momentum. Competition betweenthese sources results in thermal stratification or mixing of the pool. The interplay between the sources isdetermined by the condensation regime, steam mass flow rate and pool dimensions. Thermal stratificationis a safety issue since it limits the condensing capacity of the pool and leads to higher containmentpressures in comparison to a completely mixed pool with the same average temperature. The EffectiveHeat Source (EHS) and Effective Momentum Source (EMS) models were previously developed for predictingthe macroscopic effect of steam injection and direct contact condensation phenomena on the developmentof stratification and mixing in the pool. The models provide the effective heat and momentumsources, depending on the condensation regimes. In this work we present further development of theEHS/EMS models and their implementation in the GOTHIC code for the analysis of steam injection intocontainment drywell and venting into the wetwell through the blowdown pipes. Based on thePPOOLEX experiments performed in Lappeenranta University of Technology (LUT), correlations arederived to estimate the steam condensation regime and effective heat and momentum sources as functionsof the pool and steam injection conditions. The focus is on the low steam mass flux regimes withcomplete condensation inside the blowdown pipe or chugging. Validation of the developed methodswas carried out against the PPOOLEX MIX-04 and MIX-06 tests, which showed a very good agreementbetween experimental and simulation data on the pool temperature distribution and containmentpressure.
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18.
  • Gallego-Marcos, Ignacio, et al. (författare)
  • Thermal stratification and mixing in a Nordic BWR pressure suppression pool
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 132, s. 442-450
  • Tidskriftsartikel (refereegranskat)abstract
    • The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.
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19.
  • Ghione, Alberto, 1989, et al. (författare)
  • Uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 104, s. 28-41
  • Tidskriftsartikel (refereegranskat)abstract
    • An uncertainty and sensitivity analysis for the simulation of a station blackout scenario in the Jules Horowitz Reactor (JHR) is presented. The JHR is a new material testing reactor under construction at CEA on the Cadarache site, France. The thermal-hydraulic system code CATHARE is applied to investigate the response of the reactor system to the scenario. The uncertainty and sensitivity study was based on a statistical methodology for code uncertainty propagation, and the 'Uncertainty and Sensitivity' platform URANIE was used. Accordingly, the input uncertainties relevant to the transient, were identified, quantified, and propagated to the code output. The results show that the safety criteria are not exceeded and sufficiently large safety margins exist. In addition, the most influential input uncertainties on the safety parameters were found by making use of a sensitivity analysis.
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20.
  • Helgesson, Petter, 1986-, et al. (författare)
  • Treating model defects by fitting smoothly varying model parameters : Energy dependence in nuclear data evaluation
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 120, s. 35-47
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
21.
  • Hellesen, C., 1980-, et al. (författare)
  • Benchmark and demonstration of the CHD code for transient analysis of fast reactor systems
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 109, s. 712-719
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
22.
  • Hellesen, Carl, 1980-, et al. (författare)
  • Nuclear Spent Fuel Parameter Determination using Multivariate Analysis of Fission Product Gamma Spectra
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 110, s. 886-895
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
23.
  • Holcombe, Scott, et al. (författare)
  • A Novel gamma emission tomography instrument for enhanced fuel characterization capabilities within the OECD Halden Reactor Project
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 837-845
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
24.
  • Huang, Zheng, et al. (författare)
  • Numerical investigation on quench of an ex-vessel debris bed at prototypical scale
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 122, s. 47-61
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper is concerned with the coolability of the heap-like debris beds formed in the cavity of a Nordic-type boiling water reactor (BWR) during a postulated severe accident. A numerical simulation using the MEWA code was performed to investigate the quenching process of the ex-vessel debris bed at post-dryout condition upon its formation. To qualify the simulation tool, the MEWA code was first employed to calculate the quenching tests recently conducted on the PEARL facility. Comparisons of the simulation results with the experimental measurements show a satisfactory agreement. The simulation for the debris bed of the reactor scale shows that the heap-like debris bed flooded from the top is quenched in a multi-dimensional manner. The upper region adjacent to the centerline of the bed is the most difficult for water to reach under the top-flooding condition, and thus is subject to a higher risk of remelting. The oxidation of the residual Zr in the corium has a great impact on the coolability of the debris bed due to (i) large amount of reaction heat and the subsequent positive temperature feedback, (ii) the local accumulation of the produced H2 which may create a “steam starvation” condition and suppresses the oxidation. As possible mitigation measures of oxidation, the effects of bottom-flooding and bypass on quench were also investigated. It is predicted that the debris bed becomes more quenchable with water injected from the bottom, especially for the case with the floor partially flooded in the center. A bypass channel embedded in the center of the debris bed can also promote the quenching process by providing a preferential path for both steam escape and water inflow.
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25.
  • Huang, Zheng, et al. (författare)
  • Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 128, s. 330-340
  • Tidskriftsartikel (refereegranskat)abstract
    • In the aftermath of the Fukushima Daiichi nuclear accident, a great concern has been raised about enhancing the inherent safety of a spent fuel pool (SFP). A passive cooling system using two-phase thermosiphon loops was concerned in this paper. A RELAP5/MELCOR coupling interface was developed, aiming at simultaneously simulating the transient behaviors of the SFP (by MELCOR) and the passive cooling system (by RELAP5). First the RELAP5 model of the thermosiphon loop was qualified against an experiment of a prototypical scale. Comparisons between the experiment and predictions show a good agreement. MELCOR standalone calculations for both station blackout (SBO) and loss of coolant accident (LOCA) without the passive cooling system demonstrate severe degradation of fuel rods. In contrast, for the SBO accident, the coupling simulation shows that the passive cooling system can effectively remove the decay heat, thus keeping fuel rods intact. As for the LOCA scenario, it is more challenging for the passive cooling system due to: (i) the heat transfer power is low during the drainage of water since the natural circulation of steam is blocked by the residual water at the bottom, leading to unavoidable heat-up and oxidation of fuel cladding; (ii) the heat transfer coefficient between steam and the evaporator is very small, which consequently may require a larger heat transfer surface area. Nevertheless, the heat transfer power substantially increases after the pool is emptied and natural circulation is established. The decay heat can be removed by steam convection, thus maintaining the mechanical integrity of fuel rods and stabilizing the fuel temperature eventually. It is also observed that H 2 production is undesirably promoted because the steam supply is enhanced. However such adverse effect can be diminished by increasing the thermosiphon loops number.
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26.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Commercial thorium fuel manufacture and irradiation: Testing (Th,Pu)O-2 and (Th,U)O-2 in the "Seven-Thirty" program
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 79-86
  • Tidskriftsartikel (refereegranskat)abstract
    • Thorium based fuels are being tested in the Halden Research Reactor in Norway with the aim of producing the data necessary for licensing of these fuels in today's light water reactors. The fuel types currently under irradiation are thorium oxide fuel with plutonium as the fissile component, and uranium fuel with thorium as an additive for enhancement of thermo-mechanical and neutronic fuel properties. Fuel temperatures, rod pressures and dimensional changes are monitored on-line for quantification of thermo-mechanical behavior and fission gas release. Preliminary irradiation results show benefits in terms of lower fuel temperatures, mainly caused by improved thermal conductivity of the thorium fuels. In parallel with the irradiation, a manufacturing procedure for thorium-plutonium mixed oxide fuel is developed with the aim to manufacture industrially relevant high-quality fuel pellets for the next phase of the irradiation campaign.
  •  
27.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Irradiation testing of enhanced uranium oxide fuels
  • 2019
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 125, s. 99-106
  • Tidskriftsartikel (refereegranskat)abstract
    • Enhanced uranium oxide fuel types are being tested in the Halden Research Reactor in Norway with the aim is to assess the effect that these enhancements have on fuel performance. Fuel temperatures, rod pressures and dimensional changes are being monitored online and an extensive post-irradiation examination programme is planned. Preliminary data show that fuel centerline temperatures can be lowered by addition of ThO2 to the fuel matrix, or by incorporating Cr or SiO2-TiO2 as a network structure within the fuel. In parallel, two types of cladding coatings are tested in order to investigate their in-core properties. No abnormal behaviour has been noted during the first 100 days of irradiation.
  •  
28.
  • Insulander Björk, Klara L, 1982, et al. (författare)
  • Thorium as an additive for improved neutronic properties in boiling water reactor fuel
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 113, s. 470-475
  • Tidskriftsartikel (refereegranskat)abstract
    • This article treats the replacement of burnable absorbers with a fertile absorber in boiling water reactor fuel. The target is to improve the fuel economy while meeting the same safety demands as the currently used conventional uranium oxide (UOX) fuel. A candidate fertile absorber is Th-232, and this work investigates the impact of replacing part of the U-238 in UOX fuel with Th-232. Computer simulations have been carried out and comparisons made for fuel assemblies with fertile and burnable absorbers, loaded in the boiling water reactor Oskarshamn 3, using the tools and methods that are normally employed for reload design and safety evaluation for this reactor. The results show that power balance and shutdown margins can be improved at the cost of higher enrichment needs. Alternatively, the fuel can be designed to just fulfil the relevant safety criteria, giving slightly lower uranium needs, which may compensate for the increased enrichment costs.
  •  
29.
  • Jareteg, Klas, 1986, et al. (författare)
  • Coupled fine-mesh neutronics and thermal-hydraulics - modeling and implementation for PWR fuel assemblies
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 84, s. 244-257
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper we present a fine-mesh solver aimed at resolving in a coupled manner and at the pin cell level the neutronic and thermal-hydraulic fields. Presently, the tool considers Pressurized Water Reactor (PWR) conditions. The methods and implementation strategy are such that the coupled neutronic and thermal-hydraulic problem is formulated in a fully three-dimensional (3D) and fine mesh manner, and for steady-state situations. The solver is built on finite volume discretization schemes, matrix solvers and capabilities for parallel computing that are availablein the open source C++ library foam-extend-3.0. The angular neutron flux is determined with a multigroup discrete ordinates method (SN ), solved by a sweeping algorithm. The thermal-hydraulics is based on Computational Fluid Dynamics (CFD) models for the moderator/coolant mass, momentum, and energy equations, together with the fuel pin energy equation. The multiphysics coupling is solved by making use of an iterative algorithm, and convergence is ensured for both the separate equations and the coupled scheme. Since all the equations are implemented in the same software, all fields can be directly accessed in such a manner that external transfer and external mapping are avoided. The parallelization relies on a domain decomposition which is shared between the neutronics and the thermal-hydraulics. The latter allows to exchange the coupled data locally on each CPU, thus minimizing the data transfer. The code is tested on a quarter of a 15 × 15 PWR fuel lattice. The results show that convergence is successfully reached, and correct physical behaviors of all fields can be achieved with a reasonable computational effort.
  •  
30.
  • Kajan, Ivan, 1984, et al. (författare)
  • Impact of Ag and NOx compounds on the transport of ruthenium in the primary circuit of nuclear power plant in a severe accident
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 100, s. 9-19
  • Tidskriftsartikel (refereegranskat)abstract
    • Ruthenium is a semi-volatile element originating as a fission product in nuclear reactors that can be released in case of a severe nuclear accident. In this work, the impact of atmosphere composition on the transport of ruthenium through the primary circuit was examined. The effects of silver nanoparticles representing aerosols and NO2 gas as a product of air radiolysis were studied. Quantification of ruthenium transported both as gas and aerosol was performed. Chemical composition of ruthenium species was evaluated. The transport of gaseous ruthenium through the facility increased significantly when NO2 gas was fed into the atmosphere. When both silver aerosols and NO2 were fed into the atmosphere, the transport of ruthenium in gaseous and aerosol forms was promoted. It was concluded that the composition of atmosphere in the primary circuit will have a notable effect on the speciation of ruthenium transported into the containment building during a severe accident and thus on the potential radioactive release to the environment.
  •  
31.
  • Kitamura, Yasunori, et al. (författare)
  • Determination of prompt neutron decay constant by time-domain fluctuation analyses of detector current signals
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 120, s. 691-706
  • Tidskriftsartikel (refereegranskat)abstract
    • The conventional time-domain reactor noise techniques analyse the number of neutron detector pulse signals, so that they sometimes encounter serious difficulties owing to the count-loss effect due to the dead time of detector systems. To avoid all the difficulties coming from the count-loss effect, a novel time-domain technique was recently proposed (Pál and Pázsit, 2015). This technique analyses the auto-covariance function of continuous current signals arising from ionization chambers such as the fission chamber, so that it is inherently insensitive to the count-loss effect. In the present study, a different time-domain technique that analyses the integral values of current signals is proposed. With regard to these two techniques, the experimental conditions under which they successfully measure the subcriticality through determination of the prompt neutron decay constant are clarified.
  •  
32.
  • Li, Haipeng, et al. (författare)
  • CFD model of diabatic annular two-phase flow using the Eulerian-Lagrangian approach
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 77, s. 415-424
  • Tidskriftsartikel (refereegranskat)abstract
    • A computational fluid dynamics (CFD) model of annular two-phase flow with evaporating liquid film has been developed based on the Eulerian-Lagrangian approach, with the objective to predict the dryout occurrence. Due to the fact that the liquid film is sufficiently thin in the diabatic annular flow and at the pre-dryout conditions, it is assumed that the flow in the wall normal direction can be neglected, and the spatial gradients of the dependent variables tangential to the wall are negligible compared to those in the wall normal direction. Subsequently the transport equations of mass, momentum and energy for liquid film are integrated in the wall normal direction to obtain two-dimensional equations, with all the liquid film properties depth-averaged. The liquid film model is coupled to the gas core flow, which currently is represented using the Eulerian-Lagrangian technique. The mass, momentum and energy transfers between the liquid film, gas, and entrained droplets have been taken into account. The resultant unified model for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show favorable agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.
  •  
33.
  • Li, Hua, et al. (författare)
  • Thermal stratification and mixing in a suppression pool induced by direct steam injection
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier Ltd. - 0306-4549 .- 1873-2100. ; 111, s. 487-498
  • Tidskriftsartikel (refereegranskat)abstract
    • An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool. 
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34.
  • Mickus, Ignas, et al. (författare)
  • Optimal neutron population growth in accelerated Monte Carlo criticality calculations
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 117, s. 297-304
  • Tidskriftsartikel (refereegranskat)abstract
    • We present a source convergence acceleration method for Monte Carlo criticality calculations. The method gradually increases the neutron population size over the successive inactive as well as active criticality cycles. This helps to iterate the fission source faster at the beginning of the simulation where the source may contain large errors coming from the initial cycle; and, as the neutron population size grows over the cycles, the bias in the source gets reduced. Unlike previously suggested acceleration methods that aim at optimisation of the neutron population size, the new method does not have any significant computing overhead, and moreover it can be easily implemented into existing Monte Carlo criticality codes. The effectiveness of the method is demonstrated on a number of PWR full-core criticality calculations using a modified SERPENT 2 code.
  •  
35.
  • Pal, L., et al. (författare)
  • Fluctuations of the fission energy generated in a multiplying system
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 99, s. 116-123
  • Tidskriftsartikel (refereegranskat)abstract
    • The purpose of this work is to elaborate a master equation formalism for the evolution of the probability distribution of the cumulative energy generated by fissions in a multiplying system with delayed neutrons. The formalism accounts for the fact that the fission energy chi is also a random variable, thus the fluctuations of the total energy generated are due to both the fluctuations of the number of fissions, as well as to the fluctuations of the energy per fission. By comparing to the case where the fission energy is taken as constant, the significance of the fluctuations of the fission energy can be assessed. The first two moments of the cumulative fission energy are determined explicitly, and the time dependence of the expectation and the variance is calculated for different reactivities. As expected, the variance of the energy per fission does not play a significant role in the variance of the cumulative fission energy.
  •  
36.
  • Pazsit, Imre, 1948, et al. (författare)
  • Note on the extinction probability
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 106, s. 271-272
  • Tidskriftsartikel (övrigt vetenskapligt/konstnärligt)
  •  
37.
  • Pazsit, Imre, 1948 (författare)
  • On the concept of neutron multiplication
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 90, s. 480-481
  • Tidskriftsartikel (övrigt vetenskapligt/konstnärligt)
  •  
38.
  • Pazsit, Imre, 1948, et al. (författare)
  • The dynamic adjoint as a Green's function
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 86:Special Issue, s. 29-34
  • Tidskriftsartikel (refereegranskat)abstract
    • The concept of the dynamic adjoint was introduced by Hugo van Dam for calculating the in-core neutron noise in boiling water reactors in the mid-70's. This successful approach found numerous applications for calculating the neutron noise in both PWRs and BWRs since then. Although the advantages and disadvantages of using the direct (forward) or the adjoint (backward) approach for the calculation of the neutron noise were analysed in a number of publications, the direct relationship between the forward Green's function and the dynamic adjoint has not been discussed. On the other hand, in particle transport theory the relationship between the direct and adjoint Green's function has been discussed in detail, in which Mike Williams has had many seminal contributions. In this note we analyse the relationship between the direct Green's function and the dynamic adjoint in the spirit of Mike's work in neutron transport and radiation damage theory. The paper is closed with some personal remarks and reminiscences.
  •  
39.
  • Phung, Viet-Anh, et al. (författare)
  • Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks
  • 2018
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; , s. 461-476
  • Tidskriftsartikel (refereegranskat)abstract
    • Severe accident management strategy in Nordic boiling water reactors (BWRs) employs ex-vessel corium debris coolability. In-vessel core degradation and relocation provide initial conditions for further accident progression. Characteristics of debris bed in the reactor vessel lower plenum affect corium interactions with the vessel structures, vessel failure and melt ejection mode. Melt release from the vessel determines conditions for ex-vessel accident progression and potential threats to containment integrity, i.e. steam explosion, debris bed formation and coolability. Outcomes of core relocation depend on the interplay between (i) accident scenarios, e.g. timing and characteristics of failure and recovery of safety systems and (ii) accident phenomena. Uncertainty analysis is necessary for comprehensive risk assessment. However, computational efficiency of system analysis codes such as MELCOR is one of the big obstacles. Another problem is that code predictions can be quite sensitive to small variations of the input parameters and numerical discretization, due to the highly non-linear feedbacks and discrete thresholds employed in the code models. The goal of this work is to develop a computationally efficient surrogate model (SM) for prediction of main characteristics of corium debris in the vessel lower plenum of a Nordic BWR. Station black out scenario with a delayed power recovery is considered. The SM has been developed using artificial neural networks (ANNs). The networks were trained with a database of MELCOR solutions. The effect of the noisy data in the full model (FM) database was addressed by introducing scenario classification (grouping) according to the ranges of the output parameters. SMs using different number of scenario groups with/without weighting between predictions of different ANNs were compared. The obtained SM can be used for failure domain and failure probability analysis in the risk assessment framework for Nordic BWRs.
  •  
40.
  • Qvist, Staffan A., et al. (författare)
  • Tailoring the response of Autonomous Reactivity Control (ARC) systems
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 99, s. 383-398
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
41.
  • Qvist, Staffan, et al. (författare)
  • Design and performance of 2D and 3D-shuffled breed-and-burn cores
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 93-114
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
42.
  • Qvist, Staffan, 1986- (författare)
  • Optimization method for the design of hexagonal fuel assemblies
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 498-506
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
43.
  • Riber Marklund, Anders, et al. (författare)
  • Application of a new passive acoustic leak detection approach to recordings from the Dounreay prototype fast reactor
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 85, s. 175-182
  • Tidskriftsartikel (refereegranskat)abstract
    • A new approach for passive acoustic leak detection in sodium fast reactors without using a priori knowledge on the leak noise is introduced. The new approach is tested on recordings of argon and water injections from the Dounreay prototype fast reactor under digital mixing with two types of additional noise. It is estimated that the new approach is able to detect injection of argon into sodium in a stable background noise at signal to noise ratios between -9 and -17 dB with a low false alarm rate and with few free parameters in the signal processing. For detection of water into sodium injection the corresponding signal to noise ratios range from -9 to -18 dB.
  •  
44.
  • Riber Marklund, Anders, 1982-, et al. (författare)
  • Demonstration of an improved passive acoustic fault detection method on recordings from the Phénix steam generator operating at full power
  • 2017
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 101, s. 1-14
  • Tidskriftsartikel (refereegranskat)abstract
    • A hidden Markov model method proposed earlier for passive acoustic leak detection in sodium fast reactor systems has been improved in order to clarify how to set all free model parameters and to allow smaller amounts of training data. The method is based on training the model on known background noise only and optimizing its free model parameters by a parametric study of detection performance for synthetic noises superposed onto the same background. This means that the method is not assuming any knowledge on the noise to be detected and may be used as a general fault detection method, even if the application envisaged here is leak detection for sodium fast reactors. Using recordings of background noise as well as from argon injection tests performed at full power in the Phénix sodium fast reactor plant, it is estimated that the resulting method will detect leak-like deviations from the background noise with a detection delay of a few seconds, a false alarm rate close to 10-8 per second and at signal-to-noise ratio conditions at least corresponding to an additive signal at −10 dB. The method is one-channel, i.e. using input from one single acoustic sensor only.
  •  
45.
  • Rochman, D., et al. (författare)
  • A Bayesian Monte Carlo method for fission yield covariance information
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 95, s. 125-134
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
46.
  • Rochman, D., et al. (författare)
  • Nuclear data uncertainty for criticality-safety : Monte Carlo vs. linear perturbation
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 92, s. 150-160
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
47.
  • Rochman, D., et al. (författare)
  • Re-evaluation of the thermal neutron capture cross section of Nd-147
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 94, s. 612-617
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
48.
  • Sato, T., et al. (författare)
  • Overview of particle and heavy ion transport code system PHITS
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 82, s. 110-115
  • Tidskriftsartikel (refereegranskat)abstract
    • A general purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, is being developed through the collaboration of several institutes in Japan and Europe. The Japan Atomic Energy Agency is responsible for managing the entire project. PHITS can deal with the transport of nearly all particles, including neutrons, protons, heavy ions, photons, and electrons, over wide energy ranges using various nuclear reaction models and data libraries. It is written in Fortran language and can be executed on almost all computers. All components of PHITS such as its source, executable and data-library files are assembled in one package and then distributed to many countries via the Research Organization for Information Science and Technology, the Data Bank of the Organization for Economic Co-operation and Development's Nuclear Energy Agency, and the Radiation Safety Information Computational Center. More than 1500 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. This paper briefly summarizes the physics models implemented in PHITS, and introduces some important functions useful for specific applications, such as an event generator mode and beam transport functions. (C) 2014 Elsevier Ltd. All rights reserved.
  •  
49.
  • Sehgal, B. R., et al. (författare)
  • Severe accident progression in the BWR lower plenum and the modes of vessel failure
  • 2016
  • Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100.
  • Tidskriftsartikel (refereegranskat)abstract
    • Most of our knowledge base on the severe accident progression in the lower plenum of LWRs is based on the data obtained from the TMI-2 accident. It should be recognized that the lower plenum of a BWR is very different from that of a PWR. Unlike the PWR, the BWR plenum is full of control rod guide tubes (CRGTs) with their axial structural variations. These CRGTs are arranged in a cellular fashion with each CRGT supporting 4 rod bundles. There are also a large number of instrument guide tubes (IGTs), each generally placed in the middle of 4CRGTs. Both the CRGTs and IGTs traverse the thick vessel bottom wall and are welded to their extensions which come to bottom of the core. The core-melt progression in the lower plenum is controlled by the structures present and they, in turn, influence the timings and the modes of vessel failure for a BWR.The uranium oxide-zirconium oxide core melt formed in the 4 fuel bundles is directed by the structure below toward the water regions in-between the 4 CRGTs. The FCI will take place in those water regions and some particulate debris will be created, although there is insufficient water for quenching the melt. A FCI may occur inside a CRGT if and when the melt enters the CRGT at its top opening or the melt in the water region between the four CRGTs breaches the wall of the CRGT.The important issue is whether the welding holding the IGT inside the vessel will fail and the bottom part of the IGT falls out creating a hole in the vessel with release of water and melt/particulate debris from the vessel to the dry well of the BWR containment. Similarly, the failure of CRGT could have water and melt/particulate debris coming out of the vessel. These modes of vessel failure appear to be credible and they could occur before any large-scale melting and melt pool convection takes place. These modes of vessel failure and the melt release to the containment will have very different consequences than those generated by the other modes of vessel failure.Such BWR plenum melt progression scenarios have been considered in this paper. Some results of analyses performed at KTH have been described. We believe that the issues raised are important enough to consider a set of experiments for verification and validation of the melt progression in a BWR plenum. Such experiments are proposed.
  •  
50.
  • Suvdantsetseg, Erdenechimeg, et al. (författare)
  • Preliminary transient analysis of the Autonomous Reactivity Control system for fast reactors
  • 2015
  • Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 77, s. 47-64
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
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