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  • 1. Arzhanov, Vasily
    et al.
    Pazsit, I.
    Garis, N. S.
    Localization of a vibrating control rod pin in pressurized water reactors using the neutron flux and current noise2000In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 131, no 2, p. 239-251Article in journal (Refereed)
    Abstract [en]

    It has been proposed that the fluctuations of the neutron current called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The possibility of the localization of a vibrating control rod pin in a pressurized water reactor control assembly is investigated by using the scalar neutron noise and the two-dimensional radial current noise as measured at one central point in the assembly. Art explicit localization technique is elaborated in which the searched position is determined as the absolute minimum of a minimization function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method.

  • 2.
    Bechta, Sevostian
    et al.
    Alexandrov Scientific-Research Technology Institute, Sosnovy Bor.
    Granovsky, V.S.
    Alexandrov Research Institute of Technologies (NITI).
    Khabensky, V.B.
    A.P. Aleksandrov Research Institute of Technology.
    Krushinov, E.V.
    A.P. Aleksandrov Research Institute of Technology.
    Vitol, S.A
    A.P. Aleksandrov Research Institute of Technology.
    Sulatsky, A.A.
    Alexandrov Scientific-Research Institute of Technology.
    Gusarov, V.V.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Almjashev, V.I.
    Institute of Silicate Chemistry of Russian Academy of Sciences.
    Lopukh, D.B.
    Saint-Petersburg Electrotechnical University ‘LETI’.
    Bottomley, D.
    European Commission – DG – Joint Research Centre, Institute for Transuranium Elements.
    Fischer, M.
    Framatome ANP GmbH.
    Piluso, P.
    CEA/DEN/DSNI, Bbt. 121, 91191 Gif sur Yvette Cedex, Saclay, France.
    Miassoedov, A.
    Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe.
    Tromm, W.
    FZK, IKET, Karlsruhe, Forschungszentrum Karlsruhe GmbH.
    Altstadt, E.
    Forschungszentrum Rossendorf (FZR), 01328 Dresden, Germany.
    Fichot, F.
    Institut de Radioprotection et Sûreté Nucléaire IRSN/DPAM.
    Kymaelaeinen, O.
    Fortum Nuclear Services Ltd, POB 10, FIN-00048 Fortum (Finland).
    INTERACTION BETWEEN MOLTEN CORIUM UO2+x-ZrO2-FeOy AND VVER VESSEL STEEL2010In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 170, no 1, p. 210-218Article in journal (Refereed)
    Abstract [en]

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO2+x-ZrO2-FeOy melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction zone. The available experimental data have been used to develop a correlation for the corrosion rate as afunction of temperature and heat flux.

  • 3. Cholewa, W
    et al.
    Frid, Wiktor
    KTH, Superseded Departments, Physics.
    Bednarski, Marcin
    KTH, Superseded Departments, Physics.
    Identification of loss-of-coolant accidents in LWRs by inverse models2004In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 147, no 2, p. 216-226Article in journal (Refereed)
    Abstract [en]

    This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model.

  • 4.
    Eriksson, Marcus
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Jolkkonen, Mikael
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Cahalan, James E.
    Argonne National Laboratory, Nuclear Engineering Division.
    Inherent Safety of Fuels for Accelerator-driven Systems2005In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 151, no 3, p. 314-333Article in journal (Refereed)
    Abstract [en]

    Transient safety characteristics of accelerator-driven systems using advanced minor actinide fuels have been investigated. Results for a molybdenum-based Ceramic-Metal (CerMet) fuel, a magnesia-based Ceramic-Ceramic fuel, and a zirconium-nitride-based fuel are reported. The focus is on the inherent safety aspects of core design. Accident analyses are carried out for the response to unprotected loss-of-flow and accelerator beam-overpower transients and coolant voiding scenarios. An attempt is made to establish basic design limits for the fuel and cladding. Maximum temperatures during transients are determined and compared with design limits. Reactivity effects associated with coolant void, fuel and structural expansion, and cladding relocation are investigated. Design studies encompass variations in lattice pitch and pin diameter. Critical mass studies are performed. The studies indicate favorable inherent safety features of the CerMet fuel. Major consideration is given to the potential threat of coolant voiding in accelerator-driven design proposals. Results for a transient test case study of a postulated steam generator tube rupture event leading to extensive coolant voiding are presented. The study underlines the importance of having a low coolant void reactivity value in a lead-bismuth system despite the high boiling temperature of the coolant. It was found that the power rise following a voiding transient increases dramatically near the critical state. The studies suggest that a reactivity margin of a few dollars in the voided state is sufficient to permit significant reactivity insertions.

  • 5.
    Gajev, Ivan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xu, Yunlin
    Downar, Thomas
    Sensitivity and Uncertainty of OECD Benchmark Ringhals-1TRACE/PARCS Stability Prediction2012In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 180, no 3, p. 383-398Article in journal (Refereed)
    Abstract [en]

    Unstable behavior of boiling water reactors (BWRs) is known to occur during operation at certain power and flow conditions. This paper reports on an uncertainty study of the impact of various parameters on the prediction of the stability of the BWR within the framework of the Organisation for Economic Co-operation and Development Ringhals Unit 1 (Ringhals-1) Stability Benchmark. The time domain code TRACE/PARCS was used in the analysis. The paper is divided into two parts: a sensitivity study on numerical parameters (nodalization, time step, etc.) and an uncertainty analysis of the stability event. The sensitivity study was based on a space-time converged solution, and the most important neutronic and thermal-hydraulic parameters were identified for parameterization. The uncertainty calculation was then performed using the well-established propagation of input errors methodology. Finally, the Spearman Rank method was used to identify the most influential parameters affecting the stability of Ringhals-1.

  • 6. Gottlieb, C.
    et al.
    Arzhanov, Vasily
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Garis, N.
    Feasibility study on transient identification in nuclear power plants using support vector machines2006In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 155, no 1, p. 67-77Article in journal (Refereed)
    Abstract [en]

    Support vector machines (SVMs), a relatively new paradigm in statistical learning theory, are studied for their potential to recognize transient behavior of detector signals corresponding to various accident events at nuclear power plants (NPPs). Transient classification is a major task for any computer-aided system for recognition of various malfunctions. The ability to identify the state of operation or events occurring at an NPP is crucial so that personnel can select adequate response actions. The Modular Accident Analysis Program, version 4 (MAAP4) is a program that can be used to model various normal and abnormal events in an NPP. This study uses MAAP signals describing various loss-of-coolant accidents in boiling water reactors. The simulated sensor readings corresponding to these events have been used to train and test SVM classifiers. SVM calculations have demonstrated that they can produce classifiers with good generalization ability for our data. This in, turn indicates that SVMs show promise as classifiers for the learning problem of identifying transients.

  • 7.
    Hansson, Roberta Concilio
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Park, Hyun Sun
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dynamics and preconditioning in a single-droplet vapor explosion2009In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 167, no 1, p. 223-234Article in journal (Refereed)
    Abstract [en]

    The present study aims to develop a mechanistic understanding of the thermal-hydraulic processes in a vapor explosion, which may occur in nuclear power plants during a hypothetical severe accident, involving interactions of high-temperature corium melt and volatile coolant. Dynamics of the hot liquid (melt) droplet and the volatile liquid (coolant) were investigated in the Micro-Interactions in Steam Explosion Experiments (MISTEE) facility by performing well-controlled, externally triggered, single-droplet experiments, using a high-speed visualization system with synchronized digital cinematography and continuous X-ray radiography, called Simultaneous High-speed Acquisition of X-ray Radiography and Photography (SHARP). After an elaborate image processing, the SHARP images depict the evolution of both melt material (dispersal) and coolant (bubble dynamics) and their microscale interactions. The analysis of the data shows a deficiency in using the bubble dynamics alone to provide a consistent explanation of the energetic behavior. In contrast, the SHARP data reveal a correlation between the droplet's dynamics in the bubble's first cycle and the energetics of the subsequent explosive evaporation in the bubble's second cycle. The finding provides a basis to suggest that a so-called melt-droplet preconditioning, i.e., deformation/prefragmentation of a hot melt droplet immediately following the pressure trigger, is instrumental to the subsequent coolant entrainment, evaporation, and energetics of the resulting vapor explosion.

  • 8.
    Jeltsov, Marti
    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.
    Parametric study of sloshing effects in the primary system of an isolated lead-cooled fast reactor2015In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 190, no 1, p. 1-10Article in journal (Refereed)
    Abstract [en]

    Risks related to sloshing of liquid metal coolant due to seismic excitation need to be investigated. Sloshing effects on reactor performance include first, fluid-structure interaction and second, gas entrapment in the coolant with subsequent transport of void to the core region. While the first can hypothetically lead to structural damage or coolant spill, the second increases the risk of a reactivity insertion accident and/or local dryout of the fuel. A two-dimensional computational fluid dynamics study is carried out in order to obtain insights into the modes of sloshing depending on the parameters of seismic excitation. The applicability and performance of the numerical mesh and the Eulerian volume of fluid method used to track the free surface are evaluated by modeling a simple dam break experiment. Sloshing in the cold plenum free surface region of the European Lead-cooled SYstem (ELSY) conceptual pool-type lead-cooled fast reactor (LFR) is studied. Various sinusoidal excitations are used to imitate the seismic response at the reactor level. The goal is to identify the domain of frequencies and magnitudes of the seismic response that can lead to loads threatening the structural integrity and possible core voiding due to sloshing. A map of sloshing modes has been developed to characterize the sloshing response as a function of excitation parameters. Pressure forces on vertical walls and the lid have been calculated. Finally, insight into coolant voiding has been provided.

  • 9. Kienzler, Bernhard
    et al.
    Vejmelka, Peter
    Roemer, Juergen
    Schild, Dieter
    Jansson, Mats
    KTH, School of Chemical Science and Engineering (CHE), Chemistry, Nuclear Chemistry.
    ACTINIDE MIGRATION IN FRACTURES OF GRANITE HOST ROCK: LABORATORY AND IN SITU INVESTIGATIONS2009In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 165, no 2, p. 223-240Article in journal (Refereed)
    Abstract [en]

    Within the scope of a cooperation between Svensk Karnbranslehantering AB and Forschungszentrum Karlsruhe, Institut fur Nukleare Entsorgung, a series of actinide migration experiments were performed both in the laboratory and at the Aspo Hard Rock Laboratory in Sweden. The objectives of these experiments were to quantify, the sorption of different actinide elements in single fractures of a granite host rock and to investigate the sorption mechanisms. To guarantee the most realistic conditions-as close to nature as possible-in situ experiments were performed in the Chemlab 2 borehole probe. These migration experiments were complemented by laboratory sorption and migration studies. The latter included batch experiments with flat chips of natural material extracted from fracture surfaces to identify the mineral phases relevant to radionuclide sorption by means of autoradiography. Scanning electron microscopy analyses provided information on the composition of sorption-relevant phases and X-ray photoelectron spectroscopy of Np, Tc, and Fe distribution revealed the redox states of these elements. Important mineral phases retaining all actinides and Tc were Fe-bearing phases. From the migration experiments, elution curves of the inert tracer (HTO), Np(V), U(VI), and to a small extent of Tc(VII) were obtained. Americium (III) and plutonium(IV) were not eluted. The mechanisms influencing the migration of the elements Np, U, and Tc depended on redox reactions. It was shown by various independent methods that Np(V) was reduced to the tetravalent state on the fracture surfaces, thus resulting in a pronounced dependence of the recovery on the residence time. Technetium was also retained in the tetravalent state. Elution of natural uranium from the granite drill cores was significant and is discussed in detail.

  • 10. Kozlowski, Tomasz
    et al.
    Miller, R. M.
    Barber, D. A.
    Joo, H. G.
    Consistent comparison of the codes RELAP5/PARCS and TRAC-M/PARCS for the OECD MSLB coupled code benchmark2004In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 146, no 1, p. 15-28Article in journal (Refereed)
    Abstract [en]

    A generalized interface module was developed for coupling any thermal-hydraulic code to any spatial kinetic code. In the design used here the thermal-hydraulic and spatial kinetic codes function as independent processes and communicate using the Parallel Virtual Machine software. This approach helps maximize flexibility while minimizing modifications to the respective codes. Using this interface, the U.S. Nuclear Regulatory Commission (NRC) three-dimensional neutron kinetic code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the NRC system analysis codes RELAP5 and Modernized Transient Reactor Analysis Code (TRAC-M). Consistent comparison of code results for the Organization for Economic Cooperation and Development/Nuclear Energy Agency main steam line break benchmark problem using RELAP5/PARCS and TRAC-M/PARCS was made to assess code performance.

  • 11.
    Kozlowski, Tomasz
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Peltonen, Joanna
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    QUALIFICATION OF THE RELAP5/PARCS CODE FOR BWR STABILITY EVENTS PREDICTION2011In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 174, no 1, p. 51-63Article in journal (Refereed)
    Abstract [en]

    The present study is concerned with the capability of a coupled neutron-kinetic/thermal-hydraulic code system RELAP5/PARCS for the numerical prediction of global core stability condition and instability transients. The work is motivated by the need to assess the safety significance of a number of stability transients that trigger core instability and challenge reactor protection systems. The technical approach adopted is done both to learn from real stability events and to perform analysis of idealized well-defined transients in a real plant and core configuration. In this paper, we show that the code system can serve as a unique and powerful tool to provide a consistent and reasonably reliable prediction of stability boundary even in complex plant transients. However, the prediction quality of the instability transients, i.e., core behavior without scram namely, parameters of the limit cycle remains questionable. We identify, two main factors for future studies (two-phase flow regimes in oscillatory flow and algorithm for effective grouping of thermal-hydraulic channels) as key to enhancing the predictive capability of the existing coupled code system for boiling water reactor stability.

  • 12.
    Kudinov, Pavel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Karbojian, Aram
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Dinh, Truc-Nam
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    THE DEFOR-S EXPERIMENTAL STUDY OF DEBRIS FORMATION WITH CORIUM SIMULANT MATERIALS2010In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 170, no 1, p. 219-230Article in journal (Refereed)
    Abstract [en]

    Characteristics of corium debris beds formed in a severe core melt accident are studied in the Debris Bed Formation-Snapshot (DEFOR-S) test campaign, in which superheated binary-oxidic melts (both eutectic and non-eutectic compositions) as the corium simulants are discharged into a water pool. Water subcooling and pool depth are found to significantly influence the debris fragments' morphology and agglomeration. When particle agglomeration is absent, the tests produced debris beds with porosity of similar to 60 to 70%. This porosity is significantly higher than the similar to 40% porosity broadly used in contemporary analysis of corium debris coolability in light water reactor severe accidents. The impact of debris formation on corium coolability is further complicated by debris fragments' sharp edges, roughened surfaces, and cavities that are partially or fully encapsulated within the debris fragments. These observations are made consistently in both the DEFOR-S experiments and other tests with prototypic and simulant corium melts. Synthesis of the debris fragments from the DEFOR-S tests conducted under different melt and coolant conditions reveal trends in particle size, particle sphericity, surface roughness, sharp edges, and internal porosity as functions of water subcooling and melt composition. Qualitative analysis and discussion reaffirm the complex interplay between contributing processes (droplet interfacial instability and breakup, droplet cooling and solidification, cavity formation and solid fracture) on particle morphology and, consequently, on the characteristics of the debris beds.

  • 13.
    Li, Liangxing
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Gong, Shengjie
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Ma, Weimin
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Experimental Study of Two-Phase Flow Regime and Pressure Drop in a Particulate Bed Packed with Multidiameter Particles2012In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 177, no 1, p. 107-118Article in journal (Refereed)
    Abstract [en]

    This paper documents an experimental study on two-phase flow regimes and frictional pressure drop characteristics in a particulate (porous) bed packed with multidiameter (1.5-, 3-, and 6-mm) glass spheres. The experimental results provide new data to validate/develop hydrodynamic models for coolability analysis of debris beds formed in fuel-coolant interactions during a postulated severe accident. The POMECO-FL test facility is employed to perform the experiment, with the spheres packed in a test section of 90 mm diameter and 635 mm height. The pressure drops are measured for air/water two-phase flow through the packed bed, and flow patterns are obtained by means of visual observations. Meanwhile, local void fraction in the center of the bed is measured by a microconductive probe.The experimental results show that the frictional pressure drop of single-phase flow through the bed can be predicted by the Ergun equation, if the area mean diameter of the particles is chosen in the calculation. Given the so-determined effective particle diameter, the estimation of the Reed model for two-phase flow pressure gradient in the bed has a good agreement with the experimental data. The characteristics of the local void fraction can be used to predict flow pattern and mean void fraction. It is observed that slug flow prevails when the mean void fraction is <0.5, whereas annular flow dominates after the mean void fraction is >0.7. If the effective particle diameter is further used as an influential parameter in flow pattern identification, the observed flow regimes of two-phase flow in porous media are well predicted by the existing flow pattern map.

  • 14. Liu, J. S.
    et al.
    Neretnieks, Ivars
    KTH, Superseded Departments, Chemical Engineering and Technology.
    Effect of water radiolysis caused by dispersed radionuclides on oxidative dissolution of spent fuel in a final repository2001In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 135, no 2, p. 154-161Article in journal (Refereed)
    Abstract [en]

    When released out of a canister, the radionuclides originally incorporated in the spent fuel can still deposit radiation energy (even more efficiently) into the pore water, cause water radiolysis, and produce oxidants in the buffering material. This phenomenon is termed secondary water radiolysis. The oxidants thus produced can possibly diffuse back to oxidize the spent fuel and to increase the oxidative dissolution rare of the fuel, The effect of the secondary water radiolysis has been identified and preliminarily addressed by a mass-balance model. To explore whether the effect is significant on spent-fuel dissolution, the upper-boundary limit of the effect has been set up by considering a scenario that is very unlikely to occur. Several extreme assumptions have been made: First, the canister fails completely 10(3) yr after deposition; second, the sl,ent fuel is oxidized instantaneously; and third, the radionuclides considered are those that dominantly contribute to radiolysis between 10(3) to 10(5) yr. With these assumptions, the spent-fuel dissolution rate can be increased dramatically if 10% or more of the oxidants produced by the secondary water radiolysis diffuse back to oxidize the spent fuel. It thus indicates that the effect of the secondary water radiolysis could be significant with some extreme assumptions. With more realistic assumptions, the effect could possibly become minimal. The subject is worth further investigation.

  • 15. Liu, J. S.
    et al.
    Neretnieks, Ivars
    KTH, Superseded Departments, Chemical Engineering and Technology.
    Study of the consequences of secondary water radiolysis surrounding a defective canister2003In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 142, no 3, p. 294-305Article in journal (Refereed)
    Abstract [en]

    In the concept of deep geological disposal of spent nuclear fuel, a chemically reducing environment in the near field of a repository is favorable for retaining the radionuclides in the fuel. Water radiolysis can possibly change a reducing environment in the near field to an oxidizing environment. In this paper, the consequences of secondary water radiolysis, caused by radionuclides released from the spent nuclear fuel and dispersed in the bentonite buffer surrounding a canister, have been studied. The canister is assumed to be initially defective with a hole of a few millimeters on its wall. The small hole will considerably restrict the transport of oxidants through the canister wall and the release of radionuclides to the outside of the canister. The spent fuel dissolution is assumed to be controlled by chemical kinetics at rates extrapolated from experimental studies. Two cases are considered. In the first case it is assumed that secondary phases of radionuclides [such as amorphous Pu(OH)(4) and AmOHCO3] do not precipitate inside the canister. The model results show that a relatively large domain of the near field can be oxidized by the oxidants of secondary radiolysis. In the second case it is assumed that secondary phases of radionuclides precipitate inside the canister, and the radionuclide concentration within the canister is controlled by its respective solubility limit. The amount of radionuclides released out of the canister will then be limited by the solubility of the secondary phases. The effect of the secondary radiolysis outside the canister on the rate of spent fuel oxidation inside a defective canister will be quite limited and can be neglected for any practical purposes in this case.

  • 16.
    Liu, Longcheng
    et al.
    KTH, Superseded Departments, Chemical Engineering and Technology.
    Neretnieks, Ivars
    KTH, Superseded Departments, Chemical Engineering and Technology.
    A coupled model for oxidative dissolution of spent fuel and transport of radionuclides from an initially defective canister2001In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 135, p. 273-285Article in journal (Refereed)
    Abstract [en]

    An earlier model for oxidative dissolution of spent fuel was developed by including the release behavior of actinides from the fuel surface and the barrier effect of Zircaloy claddings. The aim here is to explore the possibility and consequences of precipitation in the water film around the fuel pellets due to solubility, and transport limitations of nuclides. The model has been applied in the performance assessment of a damaged canister under natural repository conditions, by coupling to a redox-front-based model for transport of nuclides. The simulation results identify? that the time of penetration of the canister, the size of the damage, and the initial free volume of the fuel rods are important factors that dominate the dissolution behavior of the fuel matrix and thus the transport behavior of actinides in the nearfield of a repository.

  • 17.
    Liu, Longcheng
    et al.
    KTH, Superseded Departments, Chemical Engineering and Technology.
    Neretnieks, Ivars
    KTH, Superseded Departments, Chemical Engineering and Technology.
    A reactive transport model for oxidative dissolution of spent fuel and release of nuclides within a defective canister2002In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 137, p. 228-240Article in journal (Refereed)
    Abstract [en]

    In this study, we develop a mechanism-based model to take into account most of the important processes that may influence the dissolution behavior of spent fuel and subsequently the release behavior of nuclides within a defective canister in a final repository for high-level nuclear waste. The model is, in essence, a redox-controlled reactive transport model that provides a description of the mass transport of multiple species involved in both local equilibrium and kinetically controlled reactions in the system. The complexity of the kinetics of the various redox reactions involved and the requirement of the long-term prediction, however, make numerical implementation of the fully coupled model computationally inefficient. A series of scoping calculations was performed to highlight the local characteristics and behaviors of the system, and to provide a basis for refinement of the reactive transport model. The results indicate that the rapid buildup of hydrogen within the system is mainly attributed to corrosion of the cast-iron insert that primarily occurs under anaerobic conditions, rather than to radiolysis of water. The system that is rapidly in equilibrium with 50 bar hydrogen would then keep pH constant throughout the system. In addition, simulations suggest that reduction of dissolved hexavalent uranium by ferrous iron adsorbed onto the corrosion products and by dissolved H-2 are the most important mechanisms to retard the release of uranium out of the canister. More importantly, it is found that the pseudo stationary state approximation may well be applied to the system. This greatly simplifies the numerical implementation of the reactive transport model.

  • 18.
    Liu, Longcheng
    et al.
    KTH, School of Chemical Science and Engineering (CHE), Chemical Engineering and Technology, Chemical Engineering.
    Neretnieks, Ivars
    KTH, School of Chemical Science and Engineering (CHE), Chemical Engineering and Technology, Chemical Engineering.
    Analysis of fluid flow and solute transport in a fracture intersecting a canister with variable aperture fractures and arbitrary intersection angles2005In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 150, no 2, p. 132-144Article in journal (Refereed)
    Abstract [en]

    A multitude of simulations have been made for different types of rough-walled fractures, by using FEM-LAB((R)), to evaluate the mass transfer to and from water flowing through a fracture with spatially variable apertures and with an arbitrary angle of intersection to a canister that contains spent nuclear fuel. This paper presents and discusses only the results obtained for the Gaussian fractures. The simulations suggest that the intersection angle has only a minor influence on both the volumetric and the equivalent flow rates. The standard deviation of the distribution of the volumetric flow rates of the many realizations increases with increasing roughness and spatial correlation length of the aperture field, and so does that of the equivalent flow rates. The mean of the distribution of the volumetric flow rates is determined, however, solely by the hydraulic aperture, while that of the equivalent flow rates is determined by the mechanical aperture. Based upon the analytical solutions for the parallel plate model, it has been found that the distributions of both the volumetric and the equivalent flow rates are close to the Normal. Thus, two simple expressions can be devised to quantify the stochastic properties of fluid flow and solute transport through spatially variable fractures without making detailed calculations in every fracture intersecting a deposition hole or a tunnel.

  • 19.
    Liu, Longcheng
    et al.
    KTH, Superseded Departments, Chemical Engineering and Technology.
    Neretnieks, Ivars
    KTH, Superseded Departments, Chemical Engineering and Technology.
    The effect of hydrogen on oxidative dissolution of spent fuel2002In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 138, p. 69-78Article in journal (Refereed)
    Abstract [en]

    An earlier model for the oxidative dissolution of spent fuel is further developed by including the reductive effect of H-2, which is formed by both the radiolysis of ground-water and the anoxic corrosion of the cast iron insert of the canister. The kinetics of reduction of dissolved uranium species by dissolved hydrogen is derived from a series of previously published experimental studies. The simulation results suggest that the effect of autocatalytic reduction of hexavalent uranium by hydrogen may play an important role in controlling the dissolution of the fuel matrix within a canister. Further experimental studies are required to firmly verify these findings.

  • 20.
    Marklund, Lars
    et al.
    KTH, School of Architecture and the Built Environment (ABE), Land and Water Resources Engineering.
    Wörman, Anders
    KTH, School of Architecture and the Built Environment (ABE), Land and Water Resources Engineering.
    Impact of landscape topography and quaternary overburden on the performance of a geological repository of nuclear waste2008In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 163, p. 165-179Article in journal (Refereed)
    Abstract [en]

    The topographical driving forces for groundwater on different spatial scales in several ways influence the performance of a repository for nuclear waste located at large depth in crystalline bedrock. We show that the relation between local topographical characteristics (topographical steepness and wavelengths) in the area of a repository (kilometer scale) and the large-scale (hundreds of kilometers) surroundings, together with repository depth, are the primary controls of residence time distributions and the discharge pattern of radionuclides released from an underground repository. In addition, the topography affects the groundwater flow at repository depth and, therefore, influences the long-time degradation of the repository. In the areas studied, all located in Sweden, the local topography mainly controls the groundwater flow down to a depth of ∼500 m, which is the suggested depth of the Swedish repository. The importance of the large-scale topography increases with depth but even at depth where local-scale topography is dominant, the continental-scale topography affects length and depth of flow paths as well as groundwater velocities. The impact of large-scale topography is particularly clear in areas where the steepness of local-scale landforms is relatively small. The study also shows that quaternary deposits (bedrock overburden) may have a significant impact on the overall residence times in the underground because of their hydraulic and sorption properties. This effect is fiirther enhanced by the fact that flow paths originating from repository depth generally emerge in topographical lows with relatively deep layers of quaternary deposits. The findings of this study underscore the need to consider multiscale topographical characteristics as well as bedrock overburden in assessments of radiological consequences of underground repositories.

  • 21.
    Mickus, Ignas
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Dufek, Jan
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Tuttelberg, Kaur
    PERFORMANCE OF THE EXPLICIT EULER AND PREDICTOR-CORRECTOR-BASED COUPLING SCHEMES IN MONTE CARLO BURNUP CALCULATIONS OF FAST REACTORS2015In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 191, no 2, p. 193-198Article in journal (Refereed)
    Abstract [en]

    We present a stability test of the explicit Euler and predictor-corrector based coupling schemes in Monte Carlo burnup calculations of the gas fast reactor fuel assembly. Previous studies have identified numerical instabilities of these coupling schemes in Monte Carlo burnup calculations of thermal-spectrum reactors due to spatial feedback induced neutron flux and nuclide density oscillations, where only sufficiently small time steps could guarantee acceptable precision. New results suggest that these instabilities are insignificant in fast-spectrum assembly burnup calculations, and the considered coupling schemes can therefore perform well in fast-spectrum reactor burnup calculations even with relatively large time steps. Note: Some figures in this technical note may be in color only in the electronic version.

  • 22. Nilsson, Karl-Fredrik
    et al.
    Dillström, Peter
    Andersson, Claes-Göran
    Nilsson, Fred
    KTH, School of Engineering Sciences (SCI), Solid Mechanics (Dept.).
    Andersson, Mats
    Minnebo, Philip
    Björkegren, Lars-Erik
    Erixon, Bo
    A probabilistic methodology to determine failure probabilities and acceptance criteria for the KBS-3 inserts under ice-age load conditions2008In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 163, no 1, p. 3-14Article in journal (Refereed)
    Abstract [en]

    The Swedish KBS-3 copper-cast iron canister for geological disposal of spent nuclear fuel is in an advanced stage. This paper deals with the cast iron insert that provides the mechanical strength of the canister and outlines an approach to assess the failure probabilities for manufactured canisters at large isostatic pressure (44 MPa) that could occur during future glaciations and first steps to derive acceptance criteria to ensure that failure probabilities are extremely small. The work includes a statistical test program using three inserts to determine the tensile, compression, and fracture properties. Specimens used for material characterization were also investigated by microstructural analysis to determine the microstructure and to classify and size defects. It was found that the material scatter and low ductility were caused by many defect types, but slag defects in the form of oxidation films were the most important ones. These data were then used to compute defect distributions for the probabilistic failure analysis of the KBS-3 canisters. A large number of finite element-analyses of canisters were performed at the maximum design load (44 MPa) covering distributions of material parameters and geometrical features of the canisters. The computed probabilities for fracture and plastic collapse were very low even for material data with low ductility. Two large-scale isostatic compression tests of KBS-3 mock-ups to verify safety margins are also described. The failure occurred at loads above 130 MPa in both cases, indicating a safety margin of at least a factor 3 against the maximum design load. As a result of the project, new acceptance criteria are being proposed for insert geometry and material properties, and the manufacturing process for inserts has been modified to ensure that these criteria are always fulfilled.

  • 23. Painter, Scott L.
    et al.
    Cvetkovic, Vladimir
    KTH, School of Architecture and the Built Environment (ABE), Land and Water Resources Engineering, Water Resources Engineering.
    Pensado, Osvaldo
    Time-domain random-walk algorithms for simulating radionuclide transport in fractured porous rock2008In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 163, no 1, p. 129-136Article in journal (Refereed)
    Abstract [en]

    Time-domain random-walk (TDRW) algorithms are efficient methods for simulating solute transport along one-dimensional pathways. New extensions of the TDRW algorithm accommodate decay and ingrowth of radionuclides in a decay chain and time-dependent transport velocities. Tests using equilibrium sorption and matrix diffusion retention models demonstrate that the extended TDRW algorithm is accurate and computationally efficient. When combined with stochastic simulation of transport properties, the resulting algorithm, Particle on Random Streamline Segment (PORSS), also captures the effects of random spatial variations in transport velocities, including the effects of very broad velocity distributions. When used in combination with discrete fracture network simulations, the PORSS algorithm provides an accurate and practical method for simulating radionuclide transport at the geosphere scale without invoking the advection-dispersion equation.

  • 24.
    Peltonen, Joanna
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Kozlowski, Tomasz
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Development of effective algorithm for coupled thermal-hydraulic-neutron- kinetics analysis of reactivity transient2011In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 176, no 2, p. 195-210Article in journal (Refereed)
    Abstract [en]

    Analyses of nuclear reactor safety have increasingly required the coupling of full three-dimensional neutron-kinetics (NK) core models with system transient thermal-hydraulic (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial descriptions of the reactor core. The TH code uses few, typically 5 to 20, TH channels that represent the core. The NK code uses the explicit node for each fuel assembly. Therefore, a spatial mapping of coarse grid TH and fine grid NK domain is necessary. However, improper mappings may result in the loss of valuable information, thus causing inaccurate prediction of safety parameters. The purpose of this investigation is to study the sensitivity of spatial coupling (channel refinement and spatial mapping) and develop recommendations for NK-TH mapping in the simulation of safety transients control rod drop, turbine trip, and feedwater transient combined with stability performance (minimum pump speed of recirculation pumps). The research methodology consists of a spatial coupling convergence study, as an increasing number of TH channels and different mapping schemes approach the reference case. The reference case consists of one TH channel per one fuel assembly. The comparison of results has been done under steady-state and transient conditions. The obtained results and conclusions are presented in this paper.

  • 25.
    Reisch, Frigyes
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    HIGH PRESSURE BOILING WATER REACTOR2010In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 172, no 2, p. 101-107Article in journal (Refereed)
    Abstract [en]

    Some 400 boiling water reactors (BWRs) and pressurized water reactors (PWRs) have been in operation for several decades. The presented concept, the high pressure boiling water reactor (HP-BWR), combines the best parts and omits the troublesome components of traditional BWRs and PWRs by taking into consideration the experiences gained during their operation. One of the major benefits of the HP-BWR is that safety is improved. The design utilizes gravity-operated control rods, and there is a large space for the cross-formed control rods between fuel boxes. The bottom of the reactor vessel is smooth and without penetrations. All the pipe connections to the reactor vessel are well above the top of the reactor core, and core spray is not needed. Additionally, internal circulation pumps are used. The HP-BWR concept is also environmentally friendly: Improved thermal efficiency is achieved by feeding the turbine with similar to 340 degrees C (15 MPa) steam instead of similar to 285 degrees C (7 MPa), and there is less warm water release to the recipient and less uranium consumption per produced kWh, resulting in the production of less waste. Finally, the HP-BWR is cost effective and simple, operating in direct cycle mode with no need for complicated steam generators. Moisture separators and steam dryers are placed inside the reactor vessel, and additional separators and dryers can be installed inside or outside the containment. Well-proved simple dry containment or wet containment can be used. In more than half a century, an extensive regulatory licensing experience has been built from traditional BWRs and PWRs. The HP-BWR is a developed, high-performance successor of those conventional designs. Therefore, it can be expected that licensing can be accomplished in a reasonable time. Several utilities are supporting manufacturers to study concepts for future reactors. It is likely that an application to one or more electrical power companies for financial support by a manufacturer to make a detailed feasibility study of the HP-BWR would be positively treated. This could be the next step to the implementation of the HP-BWR.

  • 26. Titarenko, Y.
    et al.
    Batyaev, V.
    Titarenko, A.
    Butko, M.
    Pavlov, K.
    Florya, S.
    Tikhonov, R.
    Boyko, P.
    Kovalenko, A.
    Sobolevsky, N.
    Anashin, V.
    Mashnik, S.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Mokhov, N.
    Rakhno, I.
    Beam dump and local shielding layout around the itep radiation test facility2009In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 168, no 2, p. 472-476Article in journal (Refereed)
    Abstract [en]

    The Radiation Test Facility (RTF) is under construction at the Institute for Theoretical and Experimental Physics to control the electronics under irradiation of particles that imitate cosmic rays (protons, carbon, aluminum, iron, tin, bismuth, and uranium). For the norms of radiation safety of personnel and users of the RTF to be observed, a local shielding and beam dump must be designed. Simulations of the dose rates around the designed shielding and beam dump are carried out in the present work.

  • 27. Titarenko, Y.
    et al.
    Batyaev, V.
    Titarenko, A.
    Butko, M.
    Pavlov, K.
    Florya, S.
    Tikhonov, R.
    Sobolevsky, N.
    Mashnik, S.
    Gudowski, Wacław
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Mokhov, N.
    Rakhno, I.
    Residual radioactive nuclide formation in 0.8-GeV proton-irradiated extended Pb target2009In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 168, no 3, p. 631-636Article in journal (Refereed)
    Abstract [en]

    This work presents results of activation-aided determination of threshold reaction rates (RRs) in 92 samples of 209Bi, natPb, 197Au, 181Ta, 169Tm, natIn, 93Nb, ,64Zn, 65Cu, 63Cu, 59Co, 19F, and 12C and in 121 samples of 27Al. All the samples are aligned with the proton beam axis inside and outside the demountable 92-cm-thick Pb target of 15-cm diameter assembled of 23 4-cm-thick discs. The samples are placed on 12 target disks to reproduce the long axis distribution of protons and neutrons. The target was exposed to an 800-MeV proton beam. The total number of protons onto the target was (6.0 ± 0.5) X1015. The RRs were determined by the direct gamma spectrometry techniques. In total, 1196 gamma spectra have been measured, and about 1500 RRs have been determined. The measured RRs were simulated by the MCNPX and SHIELD codes. A generally acceptable agreement of simulations with experimental data has been found.

  • 28. Tucek, Kamil
    et al.
    Jolkkonen, Mikael
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Studies of an accelerator-driven transuranium burner with hafnium-based inert matrix fuel2007In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 157, no 3, p. 277-298Article in journal (Refereed)
    Abstract [en]

    Neutronic and burnup characteristics of an accelerator-driven transuranium burner in a startup mode were studied. Different inert and absorbing matrices as well as lattice configurations were assessed in order to identify suitable fuel and core design configurations. Monte Carlo transport and burnup codes were used in the analyses. The lattice pin pitch was varied to optimize the source efficiency and coolant void worth while respecting key thermal and material-related design constraints posed by fuel and cladding. A HfN matrix appeared to provide a good combination of neutronic, burnup, and safety characteristics: maintaining a hard neutron spectrum, yielding acceptable coolant void reactivity and source efficiency, and alleviating the burnup reactivity swing. A conceptual design of a (TRU,Hf)N fueled, lead-bismuth eutectic-cooled accelerator-driven system was developed. Twice higher neutron fission-to absorption probabilities in Am isotopes were achieved compared to reactor designs relying on ZrN or YN inert matrix fuel. The production of higher actinides in the fuel cycle is hence limited, with a Cm fraction in the equilibrium fuel being similar to 40% lower than for cores with ZrN matrix-based fuel. The burnup reactivity swing and associated power peaking in the core are managed by an appropriate choice of cycle length (100 days) and by core enrichment zoning. A safety analysis shows that the system is protected from instant damage during unprotected beam overpower transient.

  • 29.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Eriksson, Marcus
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Neutronics of minor-actinide burning accelerator-driven systems with ceramic fuel2005In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 152, no 3, p. 367-381Article in journal (Refereed)
    Abstract [en]

    We have investigated neutronic properties of lead-bismuth-cooled accelerator-driven systems with different minor-actinide-based ceramic fuels (two composite oxides and one solid-solution nitride). Adopting a transuranic composition with 40% plutonium in the initial load, transmutation rates of higher actinides (americium and curium) equal to 265 to 285 kg/GW(thermal) -yr are obtained. The smallest reactivity swing is provided by the magnesium oxide-based cercer fuel. The cercer cores, however, exhibit large coolant void worths, which is of concern in the case of gas bubble introduction into the core. Nitride and cermet cores are more stable with respect to void formation. The poorer neutron economy of the molybdenum-based cermet makes it difficult, however, to accommodate an inert matrix volume fraction exceeding 50%, a lower limit for fabricability. Higher plutonium fraction is thus required for the cermet, which would lead to lower actinide burning rates. The nitride core yields high actinide burning rates, low void worths, and acceptable reactivity losses.

  • 30.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Suvdantsetseg, Erdenechimeg
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Fokau, Andrei
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Electra: European Lead-Cooled Training Reactor2012In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 177, no 3, p. 303-313Article in journal (Refereed)
    Abstract [en]

    The design of a low-power fast neutron reactor cooled by liquid lead (ELECTRA) is presented. Application of (Pu,Zr)N fuel permits the design of a core with very small volume and fuel column height, resulting in highly negative coolant, fuel, and structure temperature coefficients and very low channel pressure drop. Full design power of 0.5 MW(thermal) may be completely removed by natural circulation in the primary circuit, thus eliminating the need for pumps. Analysis of flow stability and performance under unprotected transients show that the suggested design is very safe, supporting its intended use for training and educational purposes.

  • 31.
    Westlén, Daniel
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Neutronic and Safety Aspects of a Gas-Cooled Subcritical Core for Minor Actinide Transmutation2005In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 154, no 1, p. 41-51Article in journal (Refereed)
    Abstract [en]

    We have designed a gas-cooled accelerator-driven system dedicated to transmutation of minor actinides. Thanks to the excellent neutron economy of the uranium-free fuel employed, the pin pitch to diameter ratio (P/D) could be increased to 1.8. The increased coolant fraction allows for decay heat removal at ambient pressure. The large coolant fraction further results in a low, pressure loss-26 kPa over the core, 35 kPa in. total. Thanks to the large P/D, the elevation of the heat exchanger necessary to remove decay heat by natural circulation is just more than I m. The absence of uranium in conjunction. with the presence of 35% (heavy atom) americium in the fuel results in a low effective delayed neutron fraction and a vanishing Doppler feedback, making subcritical operation mandatory.

  • 32. Windecker, G.
    et al.
    Anglart, Henryk
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Technology.
    Phase distribution in a BWR fuel assembly and evaluation of a multidimensional multifield model2001In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 134, no 1, p. 49-61Article in journal (Refereed)
    Abstract [en]

    The phase and mass flux distribution is analyzed in the fuel bundle of a boiling water reactor (BWR). The numerical predictions of phase distribution, obtained with the multifield two-phase flow model implemented in a computational fluid dynamics code, are compared with detailed void measurements. The present model takes into account the detailed geometry of the assembly and the spatial distribution of heat sources. The influence of spacers is modeled by introducing pressure loss and turbulence sources in the momentum and turbulence equations, respectively. The model has been applied for simulation of bubbly two-phase flow for both subcooled and saturated nucleate boiling in a seven-rod bundle and a typical BWR fuel assembly. The predictions are in good agreement with tomographic measurements performed in the FRIGG loop at Westinghouse Atom.

  • 33. Wörman, Anders
    et al.
    Dverstorp, B. A.
    Klos, R. A.
    Xu, S. L.
    Role of the bio- and geosphere interface on migration pathways for Cs-135 and ecological effects2004In: Nuclear Technology, ISSN 0029-5450, E-ISSN 1943-7471, Vol. 148, no 2, p. 194-204Article in journal (Refereed)
    Abstract [en]

    An approach is described for hydrological, geochemical, and ecological process modeling in assessing the migration pathways of radionuclides from a repository for radioactive waste in crystalline bedrock back to the surface environment where dose to individual humans can occur. The approach is based on the characterization residence times in geologic media of a unit pulse of Cs-135 released from the repository. Performance assessment modeling of geosphere transport processes generally focuses on the properties of the host rock (crystalline bedrock in this case). Our approach includes a detailed representation of the quaternary deposits that overlie the bedrock. Although water residence times in quaternary deposits can be short, geochemical reactions, predominantly sorption, can increase solute residence times significantly. Moreover, the quaternary deposits govern the pathways to terrestrial and aquatic ecosystems and are of utmost importance for the assessment of doses to individual humans.

1 - 33 of 33
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