Ändra sökning
Avgränsa sökresultatet
1 - 19 av 19
RefereraExporteraLänk till träfflistan
Permanent länk
Referera
Referensformat
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Annat format
Fler format
Språk
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Annat språk
Fler språk
Utmatningsformat
  • html
  • text
  • asciidoc
  • rtf
Träffar per sida
  • 5
  • 10
  • 20
  • 50
  • 100
  • 250
Sortering
  • Standard (Relevans)
  • Författare A-Ö
  • Författare Ö-A
  • Titel A-Ö
  • Titel Ö-A
  • Publikationstyp A-Ö
  • Publikationstyp Ö-A
  • Äldst först
  • Nyast först
  • Skapad (Äldst först)
  • Skapad (Nyast först)
  • Senast uppdaterad (Äldst först)
  • Senast uppdaterad (Nyast först)
  • Disputationsdatum (tidigaste först)
  • Disputationsdatum (senaste först)
  • Standard (Relevans)
  • Författare A-Ö
  • Författare Ö-A
  • Titel A-Ö
  • Titel Ö-A
  • Publikationstyp A-Ö
  • Publikationstyp Ö-A
  • Äldst först
  • Nyast först
  • Skapad (Äldst först)
  • Skapad (Nyast först)
  • Senast uppdaterad (Äldst först)
  • Senast uppdaterad (Nyast först)
  • Disputationsdatum (tidigaste först)
  • Disputationsdatum (senaste först)
Markera
Maxantalet träffar du kan exportera från sökgränssnittet är 250. Vid större uttag använd dig av utsökningar.
  • 1.
    Alhassan, Erwin
    et al.
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Sjöstrand, Henrik
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Helgesson, Petter
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Österlund, Michael
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Pomp, Stephan
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Koning, Arjan J.
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik. Int Atom Energy Commiss IAEA, Nucl Data Sect, Vienna, Austria.
    Rochman, Dmitri
    Paul Scherrer Inst, CH-5232 Villigen, Switzerland.
    On the use of integral experiments for uncertainty reduction of reactor macroscopic parameters within the TMC methodology2016Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 88, s. 43-52Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    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.

  • 2. Arzhanov, Vasily
    et al.
    Pazsit, I.
    Diagnostics of core barrel vibrations by in-core and ex-core neutron noise2003Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 43, nr 04-jan, s. 151-158Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Diagnostics of core-barrel vibrations has-traditionally been made by use of ex-vessel neutron detector signals. We suggest that in Addition to the ex-core noise, also the in-core noise, induced by core barrel vibrations, be also used. This would enhance the possibilities of diagnostics where the number of the ex-core detectors is not sufficient or their positions are disadvantageous for effective diagnostics, especially for shell-mode vibrations. To this order, the theory of in-core noise induced by a fluctuating core boundary has been elaborated and applied to the diagnostics of beam and shell mode vibrations. The formulas were tested on some measurements taken in the Ringhals PWRs. The results confirm the validity of the model itself, and the possibilities for enhanced diagnostics were demonstrated. A more effective use of these novel possibilities requires more in-core detectors and/or better detector positioning.

  • 3.
    Bakardjieva, S.
    et al.
    Institute of Inorganic Chemistry, Czech Acad. Sci., Rez, Czech Republic.
    Barrachin, M.
    Institut de Radioprotection et Suˆrete´ Nucle´aire (IRSN).
    Bechta, Sevostian
    A.P. Alexandrov Research Institute of Technology (NITI).
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Brissoneau, L.
    CEA, DEN, Cadarache, F-13108 St Paul lez Durance, France.
    Cheynet, B.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Fischer, E.
    Thermodata, 6 rue du Tour de l'Eau, F-38400 St Martin d'Hères, France.
    Journeau, C.
    CEA, DEN, Cadarache.
    Kiselova, M.
    Nuclear Research Institute UJV, Rez, 250 68, Czech Republic.
    Mezentseva, L.P.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Wiss, T.
    European Commission, Joint Research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, Germany.
    Improvement of the European thermodynamic database NUCLEA2010Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, nr 1, s. 84-96Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Modelling of corium behaviour during a severe accident requires knowledge of the phases present at equilibrium for a given corium composition, temperature and pressure. The thermodynamic database NUCLEA in combination with a Gibbs Energy minimizer is the European reference tool to achieve this goal. This database has been improved thanks to the analysis of bibliographical data and to EU-funded experiments performed within the SARNET network, PLINIUS as well as the ISTC CORPHAD and EVAN projects. To assess the uncertainty range associated with Energy Dispersive X-ray analyses, a round-robin exercise has been launched in which a UO2-containing corium-concrete interaction sample from VULCANO has been analyzed by three European laboratories with satisfactorily small differences.

  • 4. Buck, M.
    et al.
    Buerger, M.
    Miassoedov, A.
    Gaus-Liu, X.
    Palagin, A.
    Godin-Jacqmin, L.
    Tran, Chi Thanh
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Chudanov, V.
    The LIVE program: Results of test L1 and joint analyses on transient molten pool thermal hydraulics2010Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, nr 1, s. 46-60Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The development of a corium pool in the lower head and its behaviour is still a critical issue. This concerns, in general, the understanding of a severe accident with core melting, its course, major critical phases and timing, and the influence of these processes on the accident progression as well as, in particular, the evaluation of in-vessel melt retention by external vessel flooding as an accident mitigation strategy. Previous studies were especially related to the in-vessel retention question and often just concentrated on the quasi-steady state behaviour of a large molten pool in the lower head, considered as a bounding configuration. However, non-feasibility of the in-vessel retention concept for high power density reactors and uncertainties e.g. due to layering effects even for low or medium power reactors, turns this to be insufficient. Rather, it is essential to consider the whole evolution of the accident, including e.g. formation and growth of the in-core melt pool, characteristics of corium arrival in the lower head, and molten pool behaviour after the debris re-melting. These phenomena have a strong impact on a potential termination of a severe accident. The general objective of the LIVE program at FZK is to study these phenomena resulting from core melting experimentally in large-scale 3D geometry and in supporting separate-effects tests, with emphasis on the transient behaviour. Up to now, several tests on molten pool behaviour have been performed within the LIVE experimental program with water and with non-eutectic melts (KNO3-NaNO3) as simulant fluids. The results of these experiments, performed in nearly adiabatic and in isothermal conditions, allow a direct comparison with findings obtained earlier in other experimental programs (SIMECO, ACOPO, BALI, etc.) and will be used for the assessment of the correlations derived for the molten pool behaviour. Complementary to other international programs with real corium melts, the results of the LIVE activities also provide data for a better understanding of in-core corium pool behaviour. The experimental results are being used for the development and validation of mechanistic models for the description of molten pool behaviour, In the present paper, a range of different models is used for post-test calculations and comparative analyses. This includes simplified, but fast running models implemented in the severe accident codes ASTEC and ATHLET-CD. Further, a computational tool developed at KTH (PECM model implemented in Fluent) is applied. These calculations are complemented by analyses with the CFD code CONV (thermal hydraulics of heterogeneous, viscous and heat-generating melts) which was developed at IBRAE (Nuclear Safety Institute of Russian Academy) within the RASPLAV project and was further improved within the ISTC 2936 Project.

  • 5. Buerger, M.
    et al.
    Buck, M.
    Pohlner, G.
    Rahman, S.
    Kulenovic, R.
    Fichot, F.
    Ma, Weimin
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Miettinen, J.
    Lindholm, I.
    Atkhen, K.
    Coolability of particulate beds in severe accidents: Status and remaining uncertainties2010Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, nr 1, s. 61-75Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Particulate debris beds may form during different stages of a severe core melt accident; e.g. in the degrading hot core, due to thermal stresses during reflooding, in the lower plenum, by melt flow from the core into water in the lower head, and in the cavity by melt flow out of a failing RPV into a wet cavity. Deep water pools in the cavity are used in Nordic BWRs as an accident management measure aiming at particulate debris formation and coolability. It has been elaborated in the joint work of the European Severe Accident Research Network (SARNET) in Work Package (WP) 11.1 that coolability of particulate debris, reflooding of hot debris as well as boil-off under decay heat (long-term coolability), is strongly favoured by 2D/3D effects in beds with non-homogeneous structure and shape. Especially, water inflow from the sides and via bottom regions strongly improves coolability as compared to 1D situations with top flooding, the latter being in the past the basis of analyses on coolability. Data from experiments included in the SARNET network (DEBRIS at IKE and STYX at VTT) and earlier ones (e.g. POMECO at KTH) have been used to validate key constitutive laws in 2D codes as WABE (IKE) and ICARE/CATHARE (IRSN), especially concerning flow friction and heat transfer. Major questions concern the need of the explicit use of interfacial friction to adequately treat the various flow situations in a unified approach, as well as the adequate characterization of realistic debris composed of irregularly shaped particles of different sizes. joint work has been supported by transfer of the WABE code to KTH and VTT. Concerning realistic debris, the formation from breakup of melt jets in water is investigated in the DEFOR experiments at KTH. Present results indicate that porosities in the debris might be much higher than previously assumed, which would strongly support attainment of coolability. Calculations have been performed with IKEJET/IKEMIX describing jet breakup, mixing and settling of resulting particles. Models about debris bed formation and porosity are developed at KTH. The codes have been applied to reactor conditions for analysing the potential for coolability in the different phases of a severe accident. Calculations have been performed with WABE (MEWA) implemented in ATHLET-CD and with ICARE/ICATHARE for degraded cores and debris beds in the lower plenum, under reflooding and boil-off. Ex-vessel situations have also been analysed. Strong effects of lateral water inflow and cooling by steam in hot areas have been demonstrated. In support, some typical basic configurations have been analysed, e.g. configurations with downcomers considered as possible AM measures. Melt pool formation or coolability of particulate debris is a major issue concerning melt retention in the core and the lower head. Present conclusions from those analyses for adequate modelling in ASTEC are outlined as well as remaining uncertainties. Experimental and analysis efforts and respective continued joint actions are discussed, which are needed to reach resolution of the coolability issue.

  • 6.
    Dinh, Truc-Nam
    et al.
    KTH, Tidigare Institutioner                               , Fysik.
    Konovalikhin, M. J.
    Sehgal, B. R.
    Core melt spreading on a reactor containment floor2000Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 36, nr 4, s. 405-468Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The ex-vessel core melt spreading, cooling and stabilization is proposed for a nuclear power plant containment design. Clearly, the retention and coolability of the decay-heated core debris is very much the focal point in the proposed new and advanced designs so that, in the postulated event of a severe accident, the containment integrity is maintained and the risk of radioactivity releases is eliminated. The work reported here includes three tasks (i) to review related methodology and data base, (ii) to develop the scaling methodology and (iii) to validate the assessment methodology developed by the authors. The study is based on state-of-the-art knowledge of the melt spreading phenomenology, in particular, and, of severe accident phenomenology in general.

  • 7.
    Gudowski, Waclaw
    et al.
    KTH, Tidigare Institutioner                               , Fysik.
    Arzhanov, Vasily
    Broeders, C.
    Broeders, I.
    Cetnar, J.
    Cummings, R.
    Ericsson, M.
    Fogelberg, B.
    Gaudard, C.
    Koning, A.
    Landeyro, P.
    Magill, J.
    Pazsit, I.
    Peerani, P.
    Phlippen, P.
    Piontek, M.
    Ramstrom, E.
    Ravetto, P.
    Ritter, G.
    Shubin, Y.
    Soubiale, S.
    Toccoli, C.
    Valade, M.
    Wallenius, Janne
    KTH, Tidigare Institutioner                               , Fysik.
    Youinou, G.
    Review of the European project - Impact of Accelerator-Based Technologies on Nuclear Fission Safety (IABAT)2001Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 38, nr 1-2, s. 135-151Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The IABAT project - Impact of Accelerator Based Technologies on Nuclear Fission Safety - started in 1996 in the frame of 4(th) Framework Programme of the European Union, R&D specific programme Nuclear fission safety 1994-1998, area A.2 Exploring innovative approaches/Fuel cycle concepts, as one of the first common European activities in ADS. The project was completed October 31, 1999. The overall objective of the IABAT project has been a preliminary assessment of the potential of Accelerator-Driven Systems (ADS) for transmutation of nuclear waste and for nuclear energy production with minimum waste generation. Moreover, more specific topics related to nuclear data and code development for ADS have been studied in more detail. Four ADSs have been studied for different fuel/coolant combinations: liquid metal coolant and solid fuel, liquid metal coolant and dispersed fuel, and fast and thermal molten salt systems. Target studies comprised multiple target solutions and radiation damage problems in a target environment. In a tool development part of the project a methodology of subcriticality monitoring has been developed based on Feynman-alpha and Rossi-alpha methods. Moreover, a new Monte-Carlo burnup code taking full advantage of continuous neutron cross-section data has been developed and benchmarked. Impact on the risk from high-level waste repositories fi om radiotoxicity reduction using ADS has been assessed giving no crystal-clear benefits of ADS for repository radiotoxicity reduction but concluding some important prerequisites for effective transmutation. In proliferation studies important differences between critical reactors and ADS have been underlined and non-proliferation measures have been proposed. In assessment of accelerator technology costing models have been created that allow the circular and linear accelerator options to be compared and the effect of parameter variations examined. The calculations reported show that cyclotron systems would be more economical, due mainly to the advantage of the cost of RF power supplies. However, the accelerator community regards with skepticism the possibility of transporting and extracting more than a 10mA beam current from a 1GeV cyclotron and therefore technical factors may limit the application of cyclotrons. Finally, this review summarizes development of nuclear data in the energy region between 20 Mev and 150 MeV. Neutron and proton transport data files for Fe, Ni, Pb, Th, U-238 and Pu-239 have been created. The high-energy part of the data files consists completely of results from model calculations, which are benchmarked against the available experimental data. Although there is obviously future work left regarding fine-tuning of several parts of the data files, the representation of nuclear reaction information up to 150 MeV is already better than can be attained with intranuclear cascade codes.

  • 8.
    Helgesson, Petter
    et al.
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik. Nucl Res & Consultancy Grp NRG, Petten, Netherlands.
    Sjöstrand, Henrik
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Arjan, J. Koning
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik. Nucl Res & Consultancy Grp NRG, Petten, Netherlands.
    Rydén, Jesper
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Matematisk-datavetenskapliga sektionen, Matematiska institutionen, Tillämpad matematik och statistik.
    Rochman, Dimitri
    PSI, Villigen, Switzerland.
    Alhassan, Erwin
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Pomp, Stephan
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Combining Total Monte Carlo and Unified Monte Carlo: Bayesian nuclear data uncertainty quantification from auto-generated experimental covariances2017Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 96, s. 76-96Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The Total Monte Carlo methodology (TMC) for nuclear data (ND) uncertainty propagation has been subject to some critique because the nuclear reaction parameters are sampled from distributions which have not been rigorously determined from experimental data. In this study, it is thoroughly explained how TMC and Unified Monte Carlo-B (UMC-B) are combined to include experimental data in TMC. Random ND files are weighted with likelihood function values computed by comparing the ND files to experimental data, using experimental covariance matrices generated from information in the experimental database EXFOR and a set of simple rules. A proof that such weights give a consistent implementation of Bayes' theorem is provided. The impact of the weights is mainly studied for a set of integral systems/applications, e.g., a set of shielding fuel assemblies which shall prevent aging of the pressure vessels of the Swedish nuclear reactors Ringhals 3 and 4.

    In this implementation, the impact from the weighting is small for many of the applications. In some cases, this can be explained by the fact that the distributions used as priors are too narrow to be valid as such. Another possible explanation is that the integral systems are highly sensitive to resonance parameters, which effectively are not treated in this work. In other cases, only a very small number of files get significantly large weights, i.e., the region of interest is poorly resolved. This convergence issue can be due to the parameter distributions used as priors or model defects, for example.

    Further, some parameters used in the rules for the EXFOR interpretation have been varied. The observed impact from varying one parameter at a time is not very strong. This can partially be due to the general insensitivity to the weights seen for many applications, and there can be strong interaction effects. The automatic treatment of outliers has a quite large impact, however.

    To approach more justified ND uncertainties, the rules for the EXFOR interpretation shall be further discussed and developed, in particular the rules for rejecting outliers, and random ND files that are intended to describe prior distributions shall be generated. Further, model defects need to be treated.

  • 9.
    Jason, Hou
    et al.
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA.
    Qvist, Staffan
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Kellogg, Roger
    Argonne Natl Lab, Nucl Engn Div, 9700 S Cass Ave, Argonne, IL 60439 USA.
    Greenspan, Ehud
    Univ Calif Berkeley, Dept Nucl Engn, Berkeley, CA 94720 USA.
    3D in-core fuel management optimization for breed-and-burn reactors2016Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 88, s. 58-74Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    Breed-and-burn (B&B) reactors are a special class of fast reactors that are designed to utilize low grade fuel such as depleted uranium without fuel reprocessing. One of the most challenging practical design feasibility issues faced by B&B reactors is the high level of radiation damage their fuel cladding has to withstand in order to sustain the B&B mode of operation more than twice the maximum radiation damage cladding materials were exposed to so far in fast reactors. This study explores the possibility of reducing the minimum required peak radiation damage by employment of 3-dimensional (3D) fuel shuffling that enables a significant reduction in the peak-to-average axial burnup, that is, more uniform fuel utilization. A new conceptual design of a B&B core made of axially segmented fuel assemblies was adopted to facilitate the 3D shuffling. Also developed is a Simulated Annealing (SA) algorithm to automate the search for the optimal 3D shuffling pattern (SP). The primary objective of the SA optimization is to minimize the peak radiation damage while its secondary objective is to minimize the burnup reactivity swing, radial power peaking factor and maximum change of fuel assembly power over the cycle. Also studied is the sensitivity of the 3D shuffled core performance to the number of axially stacked subassemblies, core height and power level.

    It was found that compared with the optimal 2-dimensional (2D) shuffled core, the optimal 3D shuffled B&B core made of four 70 cm long axially stacked sub-assemblies and 12 radial shuffling batches offers a 1/3 reduction of the peak radiation damage level from 534 down to 351 displacements per atom (dpa), along with a 45% increase in the average fuel discharge burnup, and hence, the depleted uranium utilization, while satisfying all major neutronics and thermal-hydraulics design constraints. For the same peak dpa level, the 3D shuffling offers more than double the uranium utilization and the cycle length relative to 2D shuffling. The minimum peak radiation damage is increased to 360 or to 403 dpa if the core is made of, respectively, three - 70 cm or two - 140 cm long axially stacked subassemblies. Reducing the length of the subassemblies of B&B cores made of three-segment assemblies from 70 cm to 60 or 50 cm results in an increase in the peak radiation damage from 360 dpa to, respectively, 368 and 397 dpa.

  • 10. Maschek, W.
    et al.
    Chen, X.
    Delage, F.
    Femandez-Carretero, A.
    Haas, D.
    Boccaccini, C. Matzerath
    Rineiski, A.
    Smith, P.
    Sobolev, V.
    Thetford, R.
    Wallenius, Janne
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Accelerator driven systems for transmutation: Fuel development, design and safety2008Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 50, nr 2-6, s. 333-340Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    European R&D for ADS design and fuel development is driven in the 6th FP of the EU by the EUROTRANS Programme. In EUROTRANS two ADS design routes are followed, the XT-ADS and the EFIT The XT-ADS is designed to provide the experimental demonstration of transmutation. The EFIT, the European Facility for Industrial Transmutation, aims at a conceptual design of a full transmuter. A key R&D issue is the choice of an adequate fuel. Various fuel forms have been assessed and CERCER and CERMET fuels, specifically the matrices MgO and Mo, have finally been selected. Within EUROTRANS, the domain 'AFTRA' is responsible to more deeply assess the behavior of these dedicated fuels and to provide the fuel database for the EFIT. The EFIT is optimized towards: a good transmutation efficiency, high burnup, low reactivity swing, low power peaking, adequate subcriticality, reasonable beam requirements and a high safety level. In the current paper the fuels under investigation are described, including their production route and their safety limits. First core designs of CERCER and CERMET fuelled 400 MWth EFITs have been developed within AFTRA. The trends found in the design studies and first safety analyses are presented.

  • 11. Nourgaliev, R. R.
    et al.
    Dinh, Truc-Nam
    KTH, Tidigare Institutioner                               , Fysik.
    Dinh, A. T.
    Haraldsson, H. O.
    Sehgal, B. R.
    The multiphase Eulerian-Lagrangian transport (MELT-3D) approach for modeling of multiphase mixing in fragmentation processes2003Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 42, nr 2, s. 123-157Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    A new numerical approach for modeling of multiphase mixing during melt jet/droplet fragmentation process is developed. Melt or debris movements are simulated by a particle transport model in a Lagrangian formulation, while thermohydraulic conditions of the surrounding medium are obtained from solution of the Navier-Stokes and energy-conservation equations written in an Eulerian formulation. The Lagrangian and the Eulerian solutions are coupled and advanced in time, with source terms included to model the interactions between the particle and the continuum phases. The method is validated against isothermal solid-sphere, and drop fragmentation experiments. It is found that the model is capable of describing the evolution of the melt-coolant multiphase mixing process with reasonable accuracy. The method is then applied to investigate fragmentation of a continuous jet. Effects of variations in jet/coolant velocities, and of coolant thermophysical properties are analyzed, with particular emphasis on their implications for the fragmentation and mixing processes.

  • 12.
    Pusch, Roland
    et al.
    Luleå tekniska universitet, Institutionen för samhällsbyggnad och naturresurser, Geoteknologi. Drawrite AB.
    Weston, Richard
    Division of Production and Mechanical Engineering, University of Lund,.
    Retraction notice to "Superior techniques for disposal of highly radioactive waste (HLW)": [Progress in Nuclear Energy 59, (2012), 75-85]2018Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 108, s. 517-Artikel i tidskrift (Refereegranskat)
  • 13.
    Pusch, Roland
    et al.
    Luleå tekniska universitet, Institutionen för samhällsbyggnad och naturresurser, Geoteknologi.
    Weston, Richard
    Lunds universitet.
    Superior techniques for disposal of highly radioactive waste (HLW)2012Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 59, nr 8, s. 75-85Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The Swedish Nuclear Fuel and Waste Company (SKB) has recently worked out a concept, KBS-3V, for disposal of highly radioactive waste in the form of spent reactor fuel, and asked for the Government’s approval and licensing. It implies blasting of tunnels at about 400 m depth and boring of large-diameter canister deposition holes extending vertically from the tunnel floor. The rock stresses will be critically high in the construction phase and lead to failure by spalling when the heat pulse from the canisters evolves. The canisters will be surrounded by dense expansive “buffer” clay for minimizing groundwater flow around and along them but the long-term performance of either of them is not adequately proven and the placement is impractical and risky. Four major changes of the concept would make it satisfactory. One involves reorientation of the deposition holes from vertical to 45° inclination in two directions for reducing the risk of rock failure. A second is to prepare ready-made stiff units of “supercontainers” with highly compacted blocks of clay tightly surrounding the canisters for simpler and safer installation of clay blocks and canisters. A third is to surround the supercontainers by clay mud that provides the dense buffer with water from start and supports the surrounding rock when the thermal pulse begins to raise the rock stresses. A fourth is to replace the proposed smectite-rich buffer by clay with higher chemical stability and lower but sufficient expandability. A possible fifth change can be to manufacture homogeneous copper canisters of HIPOW type, which would radically reduce the risk of contamination of groundwater by released radionuclides.

  • 14.
    Qvist, Staffan
    et al.
    Uppsala universitet, Teknisk-naturvetenskapliga vetenskapsområdet, Fysiska sektionen, Institutionen för fysik och astronomi, Tillämpad kärnfysik.
    Greenspan, Ehud
    University of California.
    An Autonomous Reactivity Control system for improved fast reactor safety2014Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 77, s. 32-47Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The Autonomous Reactivity Control (ARC) system is a new safety device that can passively provide negative reactivity feedback in fast reactors that is sufficient to compensate for the positive coolant density reactivity feedback even in large low-leakage cores. The ARC system is actuated by the inherent physical property of thermal expansion, and has a very small effect on core neutronics at standard operating conditions. Additionally, the ARC system does not have an identified failure mode that can introduce positive reactivity in to the core. An ARC system can be installed in conventional fuel as- semblies by replacing a limited number of fuel rods with rods that fill a safety function, providing negative reactivity to the core in the event of coolant temperature rise above nominal. These rods are of the same outer dimensions as the fuel rods, but contain smaller-diameter inner rods that are connected to liquid-filled reservoirs at the top and bottom of the assemblies. The reservoirs are filled with two separate liquids that stay liquid and immiscible throughout the applicable temperature range of fast reactor operation. The lower reservoir contains a “neutron poison” liquid with a high neutron absorption cross-section. The upper reservoir is filled with a separate liquid with a small neutron absorption cross- section. As the temperature in the assembly increases, the liquids in the reservoirs thermally expand, effectively pushing the absorbing liquid up toward the active core region while compressing the inert gas that fills the volume above the liquid between the inner and outer tubes of the ARC rods. The ARC system can be installed, or retrofitted in to existing systems, in every fuel assembly in the core. Since ARC in- stallations in individual fuel assemblies operate independently, the system has a high level of redun- dancy. ARC-systems respond to local transients as well as core-wide accident scenarios. After actuation, the system automatically returns to its initial state as temperatures decrease, without the need for intervention by reactor operators. The ARC system concept and design considerations are described and illustrated.

  • 15.
    Sandborg, Michael
    Linköpings universitet, Institutionen för medicin och hälsa, Medicinsk radiofysik. Linköpings universitet, Centrum för medicinsk bildvetenskap och visualisering, CMIV. Linköpings universitet, Hälsouniversitetet.
    Comparison between Lucite and Water as a phantom material in medical radiology1990Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 24, nr 1-3, s. 355-364Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    In quality assurance and quality control programs in medical radiology, Lucite (i.e. plexiglas, perspex polymethyl methacrylate (PMMA)) phantoms are often used, for convenience, instead of water to imitate the patient's body. Since Lucite has higher density and lower average atomic number than water, the energy absorption and particularly photon scattering characteristics differ. A comparison is made between water, soft tissue and Lucite phantoms of the same linear dimensions with respect to photon energy absorption and scattering at photon energies between 1 and 150 keV. Analog Monte Carlo methods are used to simulate the photon transport within phantom and receptor. Charged particle equilibrium is assumed since the electron transport is not simulated (kerma approximation). It is also assumed that coherent scattering occurs independently with the atoms of the compounds. The results show that the scatter-to-primary ratio, S/P, of the energy imparted to the receptor, with the receptior located on the rear side of the phantom, is significantly larger with the Lucite than with the water or soft tissue phantom. Also, the single-event distribution of the energy imparted to the receptor by the scattered photons transmitted through the phantom, differs between Lucite and water. The fraction of incident energy imparted to the phantom is at low photon energies (below 80 keV) lower in Lucite than in water. The opposite is true at energies above 80 keV where Compton scattering predominates, and the higher density of Lucite is decisive

  • 16.
    Sehgal, B. R.
    et al.
    KTH. KTH, Stockholm, Sweden..
    Van Dorsselaere, J. P.
    Albiol, T.
    Jacquemain, D.
    Journeau, C.
    Major achievements after 4,5 years of SARNET (Severe Accident Research Network of Excellence)2010Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 52, nr 1, s. 1-1Artikel i tidskrift (Övrigt vetenskapligt)
  • 17.
    Tesinsky, Milan
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Wallenius, Janne
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Reaktorfysik.
    Impact of reflector on Doppler feedback in fast reactorsIngår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224Artikel i tidskrift (Övrigt vetenskapligt)
  • 18.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part I: Physical processes, modeling and model implementation2009Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, nr 8, s. 849-859Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    This paper, and its companion paper [Tran C.T., Dinh, IN. The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application. Progress in Nuclear Energy (companion paper), in preparation] document the development, validation and applications of a simulation platform for computationally-effective, sufficiently-accurate numerical predictions of core melt-structure-water interactions in the light water reactor lower head during a postulated severe core-melting accident. The centerpiece of this work is the Effective Convectivity Model (ECM) for description of energy splitting in a core melt pool. Built on the concept of characteristic velocities in Effective Convectivity Conductivity Model and supported by the key findings obtained from Computational Fluid Dynamics (CFD) simulations of turbulent natural convection, heat transfer and phase changes in volumetrically heated liquid pools, the ECM is refined and extended to three-dimensions and phase changes to enable simulations of melt pool formation and corium coolability in complex geometry such as a Boiling Water Reactor (BWR) lower plenum.

  • 19.
    Tran, Chi-Thanh
    et al.
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    Dinh, Truc-Nam
    KTH, Skolan för teknikvetenskap (SCI), Fysik, Kärnkraftsäkerhet.
    The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application2009Ingår i: Progress in nuclear energy (New series), ISSN 0149-1970, E-ISSN 1878-4224, Vol. 51, nr 8, s. 860-871Artikel i tidskrift (Refereegranskat)
    Abstract [en]

    The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.

1 - 19 av 19
RefereraExporteraLänk till träfflistan
Permanent länk
Referera
Referensformat
  • apa
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Annat format
Fler format
Språk
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Annat språk
Fler språk
Utmatningsformat
  • html
  • text
  • asciidoc
  • rtf