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  • 1. Arzhanov, Vasily
    Monotonicity properties of k(eff) with shape change and with nesting2002In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 29, no 2, p. 137-145Article in journal (Refereed)
    Abstract [en]

    It was found that, contrary to expectations based on physical intuition, k(eff) can both increase and decrease when changing the shape of an initially regular critical system, while preserving its volume. Physical intuition would only allow for a decrease of k(eff) when the surface/volume ratio increases. The unexpected behaviour of increasing k(eff) was found through numerical investigation. For a convincing demonstration of the possibility of the non-monotonic behaviour, a simple geometrical proof was constructed. This latter proof, in turn, is based on the assumption that k(eff) can only increase (or stay constant) in the case of nesting, i.e. when adding extra volume to a system. Since we found no formal proof of the nesting theorem for the general case, we close the paper by a simple formal proof of the monotonic behaviour of k(eff) by nesting.

  • 2. Arzhanov, Vasily
    Multi-group theory of neutron noise induced by vibrating boundaries2002In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 29, no 18, p. 2143-2158Article in journal (Refereed)
    Abstract [en]

    The paper extends the one-group analysis of the neutron noise induced by fluctuating boundaries [Ann. Nucl. Energy 27(2000)1385] to the general multi-group non-homogeneous model. The full solution is given through the Green's function of the static problem, the static flux, and a quantity describing the boundary movements. A multi-group absorber model is proposed to represent the perturbation. which turns out to be very useful, for instance, to derive the point reactor and adiabatic approximations of the neutron noise arising from the oscillating boundaries. Finally, an equivalent solution is given in terms of the adjoint function.

  • 3. Bakardjieva, S.
    et al.
    Barrachin, M.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety. Aleksandrov RIT/NITI, Russia.
    Bezdicka, P.
    Bottomley, D.
    Brissonneau, L.
    Cheynet, B.
    Dugne, O.
    Fischer, E.
    Fischer, M.
    Gusarov, V.
    Journeau, C.
    Khabensky, V.
    Kiselova, M.
    Manara, D.
    Piluso, P.
    Sheindlin, M.
    Tyrpekl, V.
    Wiss, T.
    Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET22014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 110-124Article in journal (Refereed)
    Abstract [en]

    In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating institutes are presented in this paper and give hindsight into the major processes and so give clear indications for the future work, especially in light of the Fukushima accident.

  • 4.
    Basso, Simone
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Konovalenko, Alexander
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Effectiveness of the debris bed self-leveling under severe accident conditions2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 95, p. 75-85Article in journal (Refereed)
    Abstract [en]

    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.

  • 5. Becares, V.
    et al.
    Villamarin, D.
    Fernandez-Ordonez, M.
    Gonzalez-Romero, E. M.
    Berglöf, Carl
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Bournos, V.
    Fokov, Y.
    Mazanik, S.
    Serafimovich, I.
    Evaluation of the criticality constant from Pulsed Neutron Source measurements in the Yalina-Booster subcritical assembly2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 53, p. 40-49Article in journal (Refereed)
    Abstract [en]

    The prompt decay constant method and the area-ratio (Sjostrand) method constitute the reference techniques for measuring the reactivity of a subcritical system using Pulsed Neutron Source experiments (PNS). However, different experiments have shown that in many cases it is necessary to apply corrections to the experimental results in order to take into account spectral and spatial effects. In these cases, the approach usually followed is to develop different specific correction procedures for each method. In this work we discuss the validity of prompt decay constant method and the area-ratio method in the Yalina-Booster subcritical assembly and propose a general correction procedure based on Monte Carlo simulations.

  • 6. Becares, V.
    et al.
    Villamarin, D.
    Fernandez-Ordonez, M.
    Gonzalez-Romero, E. M.
    Berglöf, Carl
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Bournos, V.
    Fokov, Y.
    Mazanik, S.
    Serafimovich, I.
    Validation of ADS reactivity monitoring techniques in the Yalina-Booster subcritical assembly2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 53, p. 331-341Article in journal (Refereed)
    Abstract [en]

    The development of a reactivity monitoring system for subcritical reactors is a major task prior to industrial scale accelerator driven system (ADS) construction. Within the 6th European Framework Program, the IP-EUROTRANS project has performed a series of experiments at the Yalina-Booster subcritical assembly located at the Joint Institute for Power and Nuclear Research (JIPNR) of the National Academy of Sciences of Belarus, using a continuous (D, T) (fusion) neutron source in pulsed and continuous mode with short interruptions (beam trips). In this paper, the implementation and results of three different monitoring techniques intended to operate with continuous neutron sources will be presented, namely the source-jerk technique, the prompt decay constant technique and the current-to-flux technique. The results will be compared with the values of the reactivity obtained using the pulsed source in PNS experiments, discussed in detail in another paper.

  • 7.
    Bechta, Sevostian
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Miassoedov, Alexei
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Journeau, Christophe
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Okamoto, Koji
    Univ Tokyo, Tokyo, Japan..
    Manara, Dario
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Bottomley, David
    Joint Res Ctr JRC Karlsruhe, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Kurata, Masaki
    JAEA CLADS Lab, Iwaki, Fukushima, Japan..
    Sehgal, Bal Raj
    Stuckert, Jun
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Steinbrueck, Martin
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Fluhrer, Beatrix
    KIT, Hermann von Helmholtz Pl 1, D-76344 Eggenstein Leopoldshafen, Germany..
    Keim, Torsten
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Fischer, Manfred
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Langrock, Gert
    AREVA NP, Paul Gossen Str 100, D-91052 Erlangen, Germany..
    Piluso, Pascal
    Commissariat Energie Atom & Energies Alternat CEA, F-13108 St Paul Les Durance, France..
    Hozer, Zoltan
    MTA EK, Budapest, Hungary..
    Kiselova, Monika
    UJV REZ As, Hlavni 130, F-25068 Husinec Rez, Czech Republic..
    Belloni, Francesco
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    Schyns, Marc
    Belgian Nucl Res Ctr SCK, Boeretang 200, B-2400 Mol, Belgium..
    On the EU-Japan roadmap for experimental research on corium behavior2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 124, p. 541-547Article in journal (Refereed)
    Abstract [en]

    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.

  • 8.
    Berglöf, Carl
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Fernández-Ordóñez, M.
    Villamarín, D.
    Bécares, V.
    González-Romero, E. M.
    Bournos, Victor
    Muñoz-Cobo, José-Luis
    Auto-correlation and variance-to-mean measurements in a subcritical core obeying multiple alpha-modes2011In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 38, no 2-3, p. 194-202Article in journal (Refereed)
    Abstract [en]

    Neutron noise measurements based on the Rossi-alpha and Feynman-alpha methodologies have been performed in a heterogeneous subcritical system. It is shown that the traditional single alpha-mode formulations of the Rossi-alpha and Feynman-alpha methods are not applicable due to the presence of higher alpha-modes. Formalisms taking into account multiple alpha-modes are applied resulting in satisfactory results. Three alpha-modes could be identified using the Rossi-alpha method, whereas only two could be obtained using the Feynman-alpha method. In the Feynman-alpha case, the possibility to obtain the fastest decaying alpha-mode was diminished due to detector dead time effects. It was found that the slowest decaying alpha-mode does not exactly correspond to the prompt decay found in pulsed neutron source measurements, which confirms the results of previous studies. Strengths and weaknesses of the multiple alpha-mode Rossi-alpha and Feynman-alpha methods observed in this study are pointed out.

  • 9.
    Bortot, Sara
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Suvdantsetseg, E.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    BELLA: a multi-point dynamics code for safety-informed design of fast reactors2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 228-235Article in journal (Refereed)
    Abstract [en]

    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.

  • 10.
    Bortot, Sara
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Suvdantsetseg, Erdenechimeg
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Wallenius, janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    BELLA: a multi-point dynamics code for simulation of fast reactorsIn: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
    Abstract [en]

    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. 

  • 11. Chikhi, N.
    et al.
    Coindreau, O.
    Li, L. X.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Taivassalo, V.
    Takasuo, E.
    Leininger, S.
    Kulenovic, R.
    Laurien, E.
    Evaluation of an effective diameter to study quenching and dry-out of complex debris bed2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 24-41Article in journal (Refereed)
    Abstract [en]

    Many of the current research works performed in the SARNET-2 WP5 deal with the study of the coolability of debris beds in case of severe nuclear power plant accidents. One of the difficulties for modeling and transposition of experimental results to the real scale and geometry of a debris bed in a reactor is the difficulty to perform experiments with debris beds that are representative for reactor situations. Therefore, many experimental programs have been performed using beds made of multi-diameter spheres or non-spherical particles to study the physical phenomena involved in debris bed coolability and to evaluate an effective diameter. This paper first establishes the ranges of porosity and particle size distribution that might be expected for in-core debris beds and ex-vessel debris beds. Then, the results of pressure drop and dry-out heat flux (DHF) measurements obtained in various experimental setups, POMECO, DEBRIS, COOLOCE/STYX and CALIDE/PRELUDE, are presented. The issues of particle size distribution and non-sphericity are also investigated. It is shown that the experimental data obtained in "simple" debris beds are relevant to describe the behavior of more complex beds. Indeed, for several configurations, it is possible to define an "effective" diameter suitable for evaluating (with the porosity) some model parameters as well as correlations for the pressure drop across the bed, the steam flow rate during quenching and the DHF.

  • 12. Dokhane, A.
    et al.
    Judd, J.
    Gajev, Ivan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zerkak, O.
    Ferroukhi, H.
    Kozlowski, T.
    Analysis of Oskarshamn-2 stability event using TRACE/SIMULATE-3K and comparison to TRACE/PARCS and SIMULATE-3K stand-alone2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 102, p. 190-199Article in journal (Refereed)
    Abstract [en]

    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.

  • 13.
    Dufek, Jan
    Chalmers University of Technology, Applied Physics.
    Building the nodal nuclear data dependences in a many-dimensional state-variable space2011In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 38, no 7, p. 1569-1577Article in journal (Refereed)
    Abstract [en]

    We present new methods for building the polynomial-regression based nodal nuclear data models. Thedata models can reflect dependences on a large number of state variables, and they can consider varioushistory effects. Suitable multivariate polynomials that approximate the nodal data dependences are identifiedefficiently in an iterative manner. The history effects are analysed using a new sampling scheme forlattice calculations where the traditional base burnup and branch calculations are replaced by a largenumber of diverse burnup histories. The total number of lattice calculations is controlled so that the datamodels are built to a required accuracy.

  • 14.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Derivation of a stable coupling scheme for Monte Carlo burnup calculations with the thermal-hydraulic feedback2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 62, p. 260-263Article in journal (Refereed)
    Abstract [en]

    Numerically stable Monte Carlo burnup calculations of nuclear fuel cycles are now possible with the previously derived Stochastic Implicit Euler method based coupling scheme. In this paper, we show that this scheme can be easily extended to include the thermal-hydraulic feedback during the Monte Carlo burnup simulations, while preserving its unconditional stability property. At each time step, the implicit solution (for the end-of-step neutron flux, fuel nuclide densities and thermal-hydraulic conditions) is calculated iteratively by the stochastic approximation; the fuel nuclide densities and thermal-hydraulic conditions are iterated simultaneously. This coupling scheme is derived as stable in theory; i.e.; its stability is not conditioned by the choice of time steps.

  • 15.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Eduard Hoogenboom, J.
    Description of a stable scheme for steady-state coupled Monte Carlo-thermal-hydraulic calculations2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 68, p. 1-3Article in journal (Refereed)
    Abstract [en]

    We provide a detailed description of a numerically stable and efficient coupling scheme for steady-state Monte Carlo neutronic calculations with thermal-hydraulic feedback. While we have previously derived and published the stochastic approximation based method for coupling the Monte Carlo criticality and thermal-hydraulic calculations, its possible implementation has not been described in a step-by-step manner. As the simple description of the coupling scheme was repeatedly requested from us, we have decided to make it available via this note.

  • 16.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    An efficient parallel computing scheme for Monte Carlo criticality calculations2009In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 36, no 8, p. 1276-1279Article in journal (Refereed)
    Abstract [en]

    The existing parallel computing schemes for Monte Carlo criticality calculations suffer from a low efficiency when applied on many processors. We suggest a new fission matrix based scheme for efficient parallel computing. The results are derived from the fission matrix that is combined from all parallel simulations. The scheme allows for a practically ideal parallel scaling as no communication among the parallel simulations is required, and inactive cycles are not needed. (C) 2009 Elsevier Ltd. All rights reserved.

  • 17.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Fission matrix based Monte Carlo criticality calculations2009In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 36, no 8, p. 1270-1275Article in journal (Refereed)
    Abstract [en]

    We have described a fission matrix based method that allows to cancel the inactive cycles in Monte Carlo criticality calculations. The fission matrix must be sampled in the course of the Monte Carlo calculation using a space mesh with sufficiently small zones as it causes the fission matrix be insensitive to errors in the initial fission source. The k(eff) and other quantities can be derived by means of the final fission matrix. The confidence interval for the k(eff) estimate can be conservatively determined via the variance in the fission matrix.

  • 18.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Stability and convergence problems of the Monte Carlo fission matrix acceleration methods2009In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 36, no 10, p. 1648-1651Article in journal (Refereed)
    Abstract [en]

    The Monte Carlo fission matrix acceleration methods aim at accelerating the convergence of the fission source in inactive cycles of Monte Carlo criticality calculations. In practice, however, these methods may corrupt the fission source, or slow down its convergence. These phenomena have not been completely understood so far. We demonstrate the convergence problems, and explain their reasons.

  • 19.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Holst, Gustaf
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Correlation of errors in the Monte Carlo fission source and the fission matrix fundamental-mode eigenvector2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 94, p. 415-421Article in journal (Refereed)
    Abstract [en]

    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.

  • 20.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Kotlyar, Dan
    Shwageraus, Eugene
    The stochastic implicit Euler method - A stable coupling scheme for Monte Carlo burnup calculations2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 60, p. 295-300Article in journal (Refereed)
    Abstract [en]

    Existing Monte Carlo burnup codes use various schemes to solve the coupled criticality and bumup equations. Previous studies have shown that the coupling schemes of the existing Monte Carlo burnup codes can be numerically unstable. Here we develop the Stochastic Implicit Euler method - a stable and efficient new coupling scheme. The implicit solution is obtained by the stochastic approximation at each time step. Our test calculations demonstrate that the Stochastic Implicit Euler method can provide an accurate solution to problems where the methods in the existing Monte Carlo burnup codes fail.

  • 21.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Kotlyar, Dan
    Shwageraus, Eugene
    Leppänen, Jaakko
    Numerical stability of the predictor-corrector method in Monte Carlo burnup calculations of critical reactors2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 56, p. 34-38Article in journal (Refereed)
    Abstract [en]

    Monte Carlo burnup codes use various schemes to solve the coupled criticality and burnup equations. Previous studies have shown that the simplest methods, such as the beginning-of-step and middle-of-step constant flux approximations, are numerically unstable in fuel cycle calculations of critical reactors. Here we show that even the predictor-corrector methods that are implemented in established Monte Carlo burnup codes can be numerically unstable in cycle calculations of large systems.

  • 22.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI).
    Tuttelberg, Kaur
    Tallinn University of Technology.
    Monte Carlo criticality calculations accelerated by a growing neutron population2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 94, p. 16-21Article in journal (Refereed)
    Abstract [en]

    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.

  • 23.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Valtavirta, Ville
    Time step length versus efficiency of Monte Carlo burnup calculations2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 72, p. 409-412Article in journal (Refereed)
    Abstract [en]

    We demonstrate that efficiency of Monte Carlo burnup calculations can be largely affected by the selected time step length. This study employs the stochastic implicit Euler based coupling scheme for Monte Carlo burnup calculations that performs a number of inner iteration steps within each time step. In a series of calculations, we vary the time step length and the number of inner iteration steps; the results suggest that Monte Carlo burnup calculations get more efficient as the time step length is reduced. More time steps must be simulated as they get shorter; however, this is more than compensated by the decrease in computing cost per time step needed for achieving a certain accuracy.

  • 24.
    Eriksson, Marcus
    et al.
    KTH, Superseded Departments, Physics.
    Cahalan, James E.
    Argonne National Laboratory, Reactor Analysis and Engineering Division.
    Inherent Shutdown Capabilities in Accelerator-driven Systems2002In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 29, no 14, p. 1689-1706Article in journal (Refereed)
    Abstract [en]

    The applicability for inherent shutdown mechanisms in accelerator-driven systems (ADS) has been investigated. We study the role of reactivity feedbacks. The benefits, in terms of dynamics performance, for enhancing the Doppler effect are examined. Given the performance characteristics of source-driven systems, it is necessary to manage the neutron source in order to achieve inherent shutdown. The shutdown system must be capable of halting the external source before excessive temperatures are obtained. We evaluate methods, based on the analysis of unprotected accidents, to accomplish such means. Pre-concepted designs for self-actuated shutdown of the external source suggested. We investigate time responses and evaluate methods to improve the performance of the safety system. It is shown that maximum beam output must be limited by fundamental means in order to protect against accident initiators that appear to be achievable in source driven systems. Utilizing an appropriate burnup control strategy plays a key role in that effort.

  • 25. Fichot, F.
    et al.
    Carénini, L.
    Sangiorgi, M.
    Hermsmeyer, S.
    Miassoedov, A.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Zdarek, J.
    Guenadou, D.
    Some considerations to improve the methodology to assess In-Vessel Retention strategy for high-power reactors2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 119, p. 36-45Article in journal (Refereed)
    Abstract [en]

    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).

  • 26.
    Fokau, Andrei
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Zhang, Youpeng
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Ishida, Shinya
    Tokyo Institute of Technology, Department of Nuclear Engineering, Sekimoto Laboratory, Japan.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    A source efficient ADS for minor actinides burning2010In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 37, no 4, p. 540-545Article in journal (Refereed)
    Abstract [en]

    Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.

  • 27.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    University of Illinois, United States .
    Sensitivity analysis of input uncertain parameters on BWR stability using TRACE/PARCS2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 67, p. 49-58Article in journal (Refereed)
    Abstract [en]

    The unstable behavior of Boiling Water Reactors (BWR), which is known to occur at certain power and flow conditions, could cause SCRAM and decrease the economic performance of the plant. For better prediction of BWR stability and understanding of influential parameters, two TRACE/PARCS models of Ringh-als-1 and Oskarshamn-2 BWRs were employed to perform a sensitivity study. Using the propagation of input errors uncertainty method's results, an attempt has been made to identify the most influential parameters affecting the stability. Furthermore, a methodology using the spearman rank correlation coefficient has been used to identify the most influential parameters on the stability parameters (decay ratio and frequency).

  • 28.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    Univ Illinois, USA.
    Space–time convergence analysis on BWR stability using TRACE/PARCS2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 51, p. 295-306Article in journal (Refereed)
    Abstract [en]

    Unstable behavior of Boiling Water Reactors (BWRs) is known to occur during operation at certain power and flow conditions. Even though BWR instability is not a severe safety concern, it could cause reactor scram and significantly decrease the economic performance of the plant. This paper aims to (a) quantify TRACE/PARCS space–time discretization error for simulation of BWR stability, (b) establish space (nodalization) and time discretization necessary for space–time converged model and (c) show that the space–time converged model gives more reliable results for both stable and unstable reactor. The space–time converged model is obtained when further refinement of numerical discretization parameters (nodalization and time step) has negligible effect on the solution. The study is significant because performing a space–time convergence analysis is a necessary step of qualification of the TRACE/PARCS model, and use of the space–time converged model increases confidence in the prediction of BWR stability.

  • 29.
    Gallego Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Modelling of pool stratification and mixing induced by steam injectionthrough blowdown pipes2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 112, p. 624-639Article in journal (Refereed)
    Abstract [en]

    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.

  • 30.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression PoolIn: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
    Abstract [en]

    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 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 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.

  • 31.
    Gallego-Marcos, Ignacio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Thermal stratification and mixing in a Nordic BWR pressure suppression pool2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 132, p. 442-450Article in journal (Refereed)
    Abstract [en]

    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.

  • 32.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Numerical investigation on quench of an ex-vessel debris bed at prototypical scale2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 47-61Article in journal (Refereed)
    Abstract [en]

    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.

  • 33.
    Huang, Zheng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Performance of a passive cooling system for spent fuel pool using two-phase thermosiphon evaluated by RELAP5/MELCOR coupling analysis2019In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 128, p. 330-340Article in journal (Refereed)
    Abstract [en]

    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.

  • 34. Isotalo, A. E.
    et al.
    Leppänen, J.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Preventing xenon oscillations in Monte Carlo burnup calculations by enforcing equilibrium xenon distribution2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 60, p. 78-85Article in journal (Refereed)
    Abstract [en]

    Existing Monte Carlo burnup codes suffer from instabilities caused by spatial xenon oscillations. These oscillations can be prevented by forcing equilibrium between the neutron flux and saturated xenon distribution. The equilibrium calculation can be integrated to Monte Carlo neutronics, which provides a simple and lightweight solution that can be used with any of the existing burnup calculation algorithms. The stabilizing effect of this approach, as well as its limitations are demonstrated using the reactor physics code Serpent.

  • 35. Klein-Hessling, W.
    et al.
    Sonnenkalb, M.
    Jacquemain, D.
    Clement, B.
    Raimond, E.
    Dimmelmeier, H.
    Azarian, G.
    Ducros, G.
    Journeau, C.
    Herranz Puebla, L. E.
    Schumm, A.
    Miassoedov, A.
    Kljenak, I.
    Pascal, G.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Guentay, S.
    Koch, M. K.
    Ivanov, I.
    Auvinen, A.
    Lindholm, I.
    Conclusions on severe accident research priorities2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 4-11Article in journal (Refereed)
    Abstract [en]

    The objectives of the SARNET network of excellence are to define and work on common research programs in the field of severe accidents in Gen. II-III nuclear power plants and to further develop common tools and methodologies for safety assessment in this area. In order to ensure that the research conducted on severe accidents is efficient and well-focused, it is necessary to periodically evaluate and rank the priorities of research. This was done at the end of 2008 by the Severe Accident Research Priority (SARP) group at the end of the SARNET project of the 6th Framework Programme of European Commission (FP6). This group has updated this work in the FP7 SARNET2 project by accounting for the recent experimental results, the remaining safety issues as e.g. highlighted by Level 2 PSA national studies and the results of the recent ASAMPSA2 FP7 project. These evaluation activities were conducted in close relation with the work performed under the auspices of international organizations like OECD or IAEA. The Fukushima-Daiichi severe accidents, which occurred while SARNET2 was running, had some effects on the prioritization and definition of new research topics. Although significant progress has been gained and simulation models (e.g. the ASTEC integral code, jointly developed by IRSN and GRS) were improved, leading to an increased confidence in the predictive capabilities for assessing the success potential of countermeasures and/or mitigation measures, most of the selected research topics in 2008 are still of high priority. But the Fukushima-Daiichi accidents underlined that research efforts had to focus still more to improve severe accident management efficiency.

  • 36. Kouhia, Virpi
    et al.
    Riikonen, Vesa
    Kauppinen, Otso-Pekka
    Purhonen, Heikki
    Austregesilo, Henrique
    Banati, Jozsef
    Cherubini, Marco
    D'Auria, Francesco
    Inkinen, Pasi
    Karppinen, Ismo
    Kral, Pavel
    Peltokorpi, Lauri
    Peltonen, Joanna
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Weber, Sebastian
    Benchmark exercise on SBLOCA experiment of PWR PACTEL facility2013In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 59, p. 149-156Article in journal (Refereed)
    Abstract [en]

    The PWR PACTEL benchmark exercise was organized in Lappeenranta, Finland by Lappeenranta University of Technology. The benchmark consisted of two phases, i.e. a blind and an open calculation task. Seven organizations from the Czech Republic, Germany, Italy, Sweden and Finland participated in the benchmark exercise, and four system codes were utilized in the benchmark simulation tasks. Two workshops were organized for launching and concluding the benchmark, the latter of which involved presentations of the calculation results as well as discussions on the related modeling issues. The chosen experiment for the benchmark was a small break loss of coolant accident experiment which was performed to study the natural circulation behavior over a continuous range of primary side coolant inventories. For the blind calculation task, the detailed facility descriptions, the measured pressure and heat losses as well as the results of a short characterizing transient were provided. For the open calculation task part, the experiment results were released. According to the simulation results, the benchmark experiment was quite challenging to model. Several improvements were found and utilized especially for the open calculation case. The issues concerned model construction, heat and pressure losses impact, interpreting measured and calculated data, non-condensable gas effect, testing several condensation and CCFL correlations, sensitivity studies, as well as break modeling. There is a clear need for user guidelines or for a collection of best practices in modeling for every code. The benchmark offered a unique opportunity to test the best practices and solutions in modeling and analyzing tasks as well as a possibility to increase knowledge about the interpretation of test results. The benchmark exercise served as a practical and rewarding forum to discuss the needs, problems and possibilities in the analysis and in producing useful data with an experiment facility. The workshops provided an advantageous site for interaction of the code users and the experimenters.

  • 37. Kozlowski, Tomasz
    et al.
    Wysocki, Aaron
    Gajev, Ivan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xu, Yunlin
    Downar, Thomas
    Ivanov, Kostadin
    Magedanz, Jeffrey
    Hardgrove, Matthew
    March-Leuba, Jose
    Hudson, Nathanael
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 67, p. 4-12Article in journal (Refereed)
    Abstract [en]

    On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations. The event was successfully modeled by the TRACE/PARCS coupled system code, and further uncertainty analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations, and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validation for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one.

  • 38.
    Li, Haipeng
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    CFD model of diabatic annular two-phase flow using the Eulerian-Lagrangian approach2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 77, p. 415-424Article in journal (Refereed)
    Abstract [en]

    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.

  • 39.
    Li, Hua
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Villanueva, Walter
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Puustinen, M.
    Laine, J.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Thermal stratification and mixing in a suppression pool induced by direct steam injection2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 111, p. 487-498Article in journal (Refereed)
    Abstract [en]

    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. 

  • 40. Meignen, Renaud
    et al.
    Raverdy, Bruno
    Buck, Michael
    Pohlner, Georg
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Brayer, Claude
    Piluso, Pascal
    Hong, Seong-Wan
    Leskovar, Matjaz
    Ursic, Mitja
    Albrecht, Giancarlo
    Lindholm, Ilona
    Ivanov, Ivan
    Status of steam explosion understanding and modelling2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 125-133Article in journal (Refereed)
    Abstract [en]

    The main results of the major international activities related to fuel coolant interactions (FCI) of the last 4-year period are presented and a summary of the knowledge gained regarding understanding and the improvements of modelling is provided. At first, the major outcomes of the OECD SERENA-2 program are presented and discussed. Important clarifications were obtained on the so-called material effect and on FCI code capabilities. We then summarise complementary analytical analyses and experimental programs performed in the frame of the SARNET community. The focus was put on the role of melt fragmentation and solidification, the impact of void on the intensity of an explosion and the triggering mechanisms. As a conclusion, tables summarising the improvements are proposed as well as research priorities.

  • 41.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Engineering.
    Optimal neutron population growth in accelerated Monte Carlo criticality calculations2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 117, p. 297-304Article in journal (Refereed)
    Abstract [en]

    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.

  • 42. Muñoz-Cobo, José-Luis
    et al.
    Berglöf, Carl
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Peña, Juan
    Villamarín, David
    Bournos, Victor
    Feynman-alpha and Rossi-alpha formulas with spatial and modal effects2011In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 38, no 2-3, p. 590-600Article in journal (Refereed)
    Abstract [en]

    Feynman-alpha and Rossi-alpha formulas including multiple alpha-modes are derived for stochastic and continuous neutron sources. The presented formalism is further developed to achieve spatial correction factors for the single alpha-mode point kinetics representations of the Feynman-alpha and Rossi-alpha formulas. As a natural extension of the multiple alpha-mode formalism, delayed neutrons are included in the Feynman-alpha formula. The obtained formulas are validated experimentally in a strongly heterogeneous system obeying multiple alpha-modes, resulting in good agreement with the presented theoretical framework.

  • 43. Pazsit, I.
    et al.
    Arzhanov, Vasily
    Linear reactor kinetics and neutron noise in systems with fluctuating boundaries2000In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 27, no 15, p. 1385-1398Article in journal (Refereed)
    Abstract [en]

    The general theory of linear reactor kinetics and that of the induced neutron noise is developed for systems with varying size, i.e. in which the position of the boundary fluctuates around a stationary value. The point kinetic and adiabatic approximations are defined by a generalisation of the flux factorisation, and the full solution of the general problem with an arbitrarily fluctuating boundary is given by the Green's function technique. The correctness of the general solution is proven both generally and also by considering the simple case of a 2-D cylindrical reactor with a fluctuating radius, in which case a direct compact solution is possible.

  • 44.
    Peltonen, Joanna
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    Effective spatial mapping for coupled code analysis of thermal-hydraulics/neutron-kinetics of boiling water reactors2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 63, p. 461-485Article, review/survey (Refereed)
    Abstract [en]

    Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. In order to produce results within a reasonable computing time, the coupled codes use two different spatial description of the reactor core. The TH code uses few, typically 5-20 TH channels, to represent the core. The NK code uses one explicit node for each fuel assembly. Therefore, a spatial mapping of a coarse grid TH and a fine grid NK domain is necessary. However, improper mappings may result in loss of valuable information, thus causing inaccurate prediction of safety parameters. In this article the study of the effectiveness of spatial coupling (channel -refinement and spatial mapping) and developed recommendations for NK/TH mapping are presented. The sensitivity of stability (measured by Decay Ratio and Frequency) to the different types of mapping schemes is analyzed against OECD/NEA Ringhals-1 Stability Benchmark data. Additionally, to increase the efficiency and applicability of spatial mapping convergence, a new mapping methodology is proposed. The new mapping approach is based on hierarchical clustering method; the method of unsupervised learning that is adopted in many different scientific fields, thanks to its flexibility and robustness. The proposed new mapping method is shown to be very successful for spatial coupling problem and can be fully automated allowing for significant time reduction in input preparation and mapping convergence study.

  • 45.
    Persson, Carl-Magnus
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Fokau, Andrei
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Serafimovich, Ivan
    Joint Institute for Power and Nuclear Research, National Academy of Science of Belarus, Minsk, Belarus.
    Bournos, Victor
    Joint Institute for Power and Nuclear Research, National Academy of Science of Belarus, Minsk, Belarus.
    Fokov, Yurii
    Joint Institute for Power and Nuclear Research, National Academy of Science of Belarus, Minsk, Belarus.
    Routkovskaia, Christina
    Joint Institute for Power and Nuclear Research, National Academy of Science of Belarus, Minsk, Belarus.
    Kiyavitskaya, Hanna
    Joint Institute for Power and Nuclear Research, National Academy of Science of Belarus, Minsk, Belarus.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Physics.
    Pulsed neutron source measurements in the subcritical ADS experiment YALINA-Booster2008In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 35, no 12, p. 2357-2364Article in journal (Refereed)
    Abstract [en]

    A subcritical zero-power source-driven coupled core, the YALINA-Booster. has been constructed for experimental investigations of neutron kinetics of source-driven systems. In this study, the reactivity of two subcritical configurations has been determined by the area ratio method. The prompt neutron decay constants have been evaluated through slope fitting of the prompt neutron decay as well as through the pulsed Rossi-alpha method. It is shown that the slope fitting method and the pulsed Rossi-alpha method give stable results whereas the area ratio method results show spatial dependence. The reasons for the spatial spread are addressed.

  • 46.
    Phung, Viet-Anh
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Grishchenko, Dmitry
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Galushin, Sergey
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Prediction of In-Vessel Debris Bed Properties in BWR Severe Accident Scenarios using MELCOR and Neural Networks2018In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, p. 461-476Article in journal (Refereed)
    Abstract [en]

    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.

  • 47. Pohlner, G.
    et al.
    Buck, M.
    Meignen, R.
    Kudinov, Pavel
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Polidoro, F.
    Takasuo, E.
    Analyses on ex-vessel debris formation and coolability in SARNET frame2014In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 74, p. 50-57Article in journal (Refereed)
    Abstract [en]

    The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the view on possible key aspects of future activities.

  • 48.
    Riber Marklund, Anders
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Michel, Frederic
    Demonstration of an improved passive acoustic fault detection method on recordings from the Phénix steam generator operating at full power2017In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 101Article in journal (Refereed)
    Abstract [en]

    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.

  • 49.
    Riber Marklund, Anders
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology. CEA CAD/DEN/DTN/STCP/LIET, France.
    Michel, F.
    Application of a new passive acoustic leak detection approach to recordings from the Dounreay prototype fast reactor2015In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 175-182Article in journal (Refereed)
    Abstract [en]

    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.

  • 50.
    Sehgal, B. R.
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Bechta, Sevostian
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Severe accident progression in the BWR lower plenum and the modes of vessel failure2016In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100Article in journal (Refereed)
    Abstract [en]

    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.

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