Change search
Refine search result
1 - 17 of 17
CiteExportLink to result list
Permanent link
Cite
Citation style
  • apa
  • harvard1
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf
Rows per page
  • 5
  • 10
  • 20
  • 50
  • 100
  • 250
Sort
  • Standard (Relevance)
  • Author A-Ö
  • Author Ö-A
  • Title A-Ö
  • Title Ö-A
  • Publication type A-Ö
  • Publication type Ö-A
  • Issued (Oldest first)
  • Issued (Newest first)
  • Created (Oldest first)
  • Created (Newest first)
  • Last updated (Oldest first)
  • Last updated (Newest first)
  • Disputation date (earliest first)
  • Disputation date (latest first)
  • Standard (Relevance)
  • Author A-Ö
  • Author Ö-A
  • Title A-Ö
  • Title Ö-A
  • Publication type A-Ö
  • Publication type Ö-A
  • Issued (Oldest first)
  • Issued (Newest first)
  • Created (Oldest first)
  • Created (Newest first)
  • Last updated (Oldest first)
  • Last updated (Newest first)
  • Disputation date (earliest first)
  • Disputation date (latest first)
Select
The maximal number of hits you can export is 250. When you want to export more records please use the Create feeds function.
  • 1. Anglart, Henryk
    et al.
    Podowski, M. Z.
    Fluid mechanics of Taylor bubbles and slug flows in vertical channels2002In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 140, no 2, p. 165-171Article in journal (Refereed)
    Abstract [en]

    Fluid mechanics of Taylor bubbles and slug flows is investigated in vertical, circular channels using detailed, three-dimensional computational fluid dynamics simulations. The Volume of Fluid model with the interface-sharpening algorithm, implemented in the commercial CFX4 code, is used to predict the shape and velocity of Taylor bubbles moving along a vertical channel. Several cases are investigated, including both a single Taylor bubble and a train of bubbles rising in water. It is shown that the potential flow solution underpredicts the water film thickness around Taylor bubbles. Furthermore, the computer simulations that are performed reveal the importance of properly modeling the three-dimensional nature of phenomena governing the motion of Taylor bubbles. Based on the present results, a new formula for the evaluation of bubble shape is derived. Both the shape of Taylor bubbles and the bubble rise velocity predicted by the proposed model agree well with experimental observations. Furthermore, the present model shows good promise in predicting the coalescence of Taylor bubbles.

  • 2.
    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
    Serafimovich, Ivan
    Mazanik, Sergei
    Fokov, Yurii
    Spatial and Source Multiplication Effects on the Area Ratio Reactivity Determination Method in a Strongly Heterogeneous Subcritical System2010In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 166, no 2, p. 134-144Article in journal (Refereed)
    Abstract [en]

    The area ratio method of Sjostrand is generally considered one of the most reliable reactivity determination methods and thus is a major candidate for off-line calibration purposes in future accelerator-driven systems for high-level waste incineration. In this work, the Sjostrand area ratio method has been evaluated experimentally under thorough conditions in the strongly heterogeneous subcritical facility YALINA-Booster. Both strengths and weaknesses of the method have been identified. Most surprisingly, it has been found that the area ratio reactivity estimates may differ a factor of 2 depending on detector position. It is also shown that this strong spatial dependence can be explained based on a simple two-region point-kinetics model and corrected by means of correction factors obtained through Monte Carlo simulations. A new Monte Carlo correction method is proposed that includes, at the same time, the spatial disturbance and the effective delayed neutron fraction. In that way, the value of the effective multiplication factor is obtained from the measured dollar reactivity without the need of calculating the effective delayed neutron fraction explicitly, and thereby, the delayed neutron transport is performed only once. Further, it has been found that the Sjostrand area ratio method is not sensitive to perturbations of the source multiplication factor.

  • 3.
    Cvetkovic, Vladimir
    et al.
    KTH, Superseded Departments, Land and Water Resources Engineering.
    Painter, S.
    Selroos, J. O.
    Comparative measures of radionuclide containment in the crystalline geosphere2002In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 142, no 3, p. 292-304Article in journal (Refereed)
    Abstract [en]

    A probabilistic model for assessing the capacity of a fractured crystalline rock volume to contain radionuclides is developed The rock volume is viewed as a network of discrete fractures through which radionuclides are transported by flowing water. Diffusive mass transfer between the open fractures and the stagnant water in the pore space of the rock matrix allow radionuclides access to mineral grains where physical and chemical processes-collectively known as sorption-can retain radionuclides. A stochastic Lagrangian framework is adopted to compute the probability that a radionuclide particle will be retained by the rock, i.e., the probability that it will decay before being released from the rock volume. A dimensionless quantity referred to as the containment index is related to this probability and proposed as a suitable measure for comparing different rock volumes; such a comparative measure may be needed, for example, in a site selection program for geological radioactive waste disposal. The probabilistic solution of the transport problem is based on the statistics of two Lagrangian variables: T, the travel time of an imaginary tracer moving with the flowing water, and beta, a suitably normalized surface area available for retention. Statistics of tau and beta may be computed numerically using site-specific discrete fracture MP network simulations. Fracture data from the well-characterized Aspo Hard Rock Laboratory site in southern Sweden are used to illustrate the implementation of the proposed containment index for six radionuclides (Sn-126, I-129, Cs-135, Np-237, Pu-239, and Se-79). It is found that fractures of small aperture imply prolonged travel times and hence long tails in both beta and tau. This, in turn, enhances retention and is favorable from a safely assessment perspective.

  • 4.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors2006In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 152, p. 274-283Article in journal (Refereed)
    Abstract [en]

    A new adaptive stochastic approximation method for an efficient Monte Carlo calculation of steady-state conditions in thermal reactor cores is described The core conditions that we consider are spatial distributions of power, neutron flux, coolant density, and strongly absorbing fission products like Xe-135. These distributions relate to each other; thus, the steady-state conditions are described by a system of nonlinear equations. When a Monte Carlo method is used to evaluate the power or neutron flux, then the task turns to a nonlinear stochastic root-finding problem that is usually solved in the iterative manner by stochastic optimization methods. One of those methods is stochastic approximation where efficiency depends on a sequence of stepsize and sample size parameters. The stepsize generation is often based on the well-known Robbins-Monro algorithm; however, the efficient generation of the sample size (number of neutrons simulated at each iteration step) was not published yet. The proposed method controls both the stepsize and the sample size in an efficient way; according to the results, the method reaches the highest possible convergence rate.

  • 5.
    Dufek, Jan
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Hoogenboom, Eduard
    Numerical Stability of Existing Monte Carlo Burnup Codes in Cycle Calculations of Critical Reactors2009In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 162, no 3, p. 307-311Article in journal (Refereed)
    Abstract [en]

    We show that major existing Monte Carlo burnup codes are numerically unstable in cycle calculations of critical reactors; spatial oscillations of the neutron flux can be observed even when relatively small time steps are used. This is caused by using the explicit Euler or midpoint method that appear to be numerically unstable with the step sizes common in cycle calculations. More stable methods that are common in deterministic burnup calculations, like the modified Euler method, can easily be introduced into the Monte Carlo burnup codes.

  • 6.
    Eriksson, Marcus
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Cahalan, James E.
    Argonne National Laboratory, Nuclear Engineering Division.
    Yang, Won Sik
    Argonne National Laboratory, Nuclear Engineering Division.
    On the Performance of Point Kinetics for the Analysis Accelerator-driven Systems2005In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 149, no 3, p. 298-311Article in journal (Refereed)
    Abstract [en]

    The ability of point kinetics to describe dynamic processes in accelerator-driven systems (ADSs) is investigated. Full three-dimensional energy-space-time-dependent calculations, coupled with thermal and hydraulic feedback effects, are performed and used as a standard of comparison. Various transient accident sequences are studied. Calculations are performed in the range of k(eff) = 0.9594 to 0.9987 to provide insight into the dependence of the performance on the subcritical level. Numerical experiments are carried out on a minor-actinide-loaded and lead-bismuth-cooled ADS. It is shown that the point kinetics approximation is capable of providing highly accurate calculations in such systems. The results suggest better precision at lower k(eff) levels. It is found that subcritical operation provides features that are favorable from a point kinetics view of application. For example, reduced sensitivity to system reactivity perturbations effectively mitigates any spatial distortions. If a subcritical reactor is subject to a change in the strength of the external source, or a change in reactivity within the subcritical range, the neutron population will adjust to a new stationary level. Therefore, within the normal range of operation, the power predicted by the point kinetics method and the associated error in comparison with the exact solution tends to approach an essentially bounded value. It was found that the point kinetics model is likely to underestimate the power rise following a positive reactivity insertion in an ADS, which is similar to the behavior in critical systems. However, the effect is characteristically lowered in subcritical versus critical or near-critical reactor operation.

  • 7. Ignatyuk, A. V.
    et al.
    Lunev, V. P.
    Shubin, Y. N.
    Gai, E. V.
    Titarenko, N. N.
    Gudowski, Waclaw
    KTH, Superseded Departments, Physics.
    Neutron and proton cross-section evaluations for Th-232 up to 150 MeV2002In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 142, no 2, p. 177-194Article in journal (Refereed)
    Abstract [en]

    Investigations aimed at the development of neutron and proton cross-section evaluations for Th-232 at intermediate energies in the range of 0 to 200 MeV are described The coupled-channel optical model has been used to calculate the neutron total, elastic, and reaction cross sections and the elastic scattering angular distributions. Evaluations of the neutron and charged particle emission cross sections and of the fission cross sections have been obtained on the basis of the statistical description that includes direct, preequilibrium, and equilibrium mechanisms of nuclear reactions. The Kalbach parameterization of angular distributions has been used to describe the double-differential cross sections of emitted neutrons and charged particles in ENDF/B-VI format.

  • 8. Ignatyuk, A. V.
    et al.
    Lunev, V. P.
    Shubin, Y. N.
    Gai, E. V.
    Titarenko, N. N.
    Ventura, A.
    Gudowski, Waclaw
    KTH, Superseded Departments, Physics.
    Neutron cross-section evaluations for U-238 up to 150 MeV2000In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 136, no 3, p. 340-356Article in journal (Refereed)
    Abstract [en]

    Investigations aimed at the development of neutron cross-section evaluations for U-238 at intermediate energies are briefly described. The coupled-channels optical model is used to calculate the neutron total, the elastic and reaction cross sections, and the elastic-scattering angular distributions. Evaluations of the neutron and charged particle emission cross sections and of the fission cross sections are obtained on the basis of the statistical description that includes direct, preequilibrium, and equilibrium mechanisms of nuclear reactions. The Kalbach parameterization of angular distributions is used to describe the double-differential cross sections of emitted neutrons and charged particles in ENDF/B-VI format.

  • 9.
    Kozlowski, Tomasz
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Xu, Yunlin
    Downar, Thomas J.
    Lee, Deokjung
    Cell Homogenization Method for Pin-by-Pin Neutron Transport Calculations2011In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 169, no 1, p. 1-18Article in journal (Refereed)
    Abstract [en]

    For practical reactor core applications, low-order transport approximations such as SP(3) have been shown to provide sufficient accuracy for both static and transient calculations with considerably less computational expense than the discrete ordinate or the full spherical harmonics methods. These methods have been applied in several core simulators where homogenization was performed at the level of the pin cell. One of the principal problems has been to recover the error introduced by pin cell homogenization. One of the basic approaches to treat pin cell homogenization error is pin cell discontinuity factors (CDFs) based on well-established generalized equivalence theory to generate appropriate group constants. The method is able to treat all sources of error together, allowing even a few-group diffusion solution with one mesh per cell to reproduce a higher-order reference solution. However, a CDF has to be derived separately for each space-angle approximation. An additional difficulty is that in practice the CDFs have to be derived from a lattice calculation from which only the scalar flux and current are available, and therefore recovery of the exact SP(N) angular moment is not possible. This paper focuses on the pin cell scale homogenization. It demonstrates derivation of the CDF for the SP(3) transport method with finite-difference spatial discretization with the limitation of only the scalar flux and interface current being available from the heterogeneous reference. The method is demonstrated using a sample benchmark application.

  • 10. Pazsit, Imre
    et al.
    Montalvo, Cristina
    Nylen, Henrik
    Andersson, Tell
    Hernandez Solis, Augusto
    KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.
    Cartemo, Petty Bernitt
    Developments in Core-Barrel Motion Monitoring and Applications to the Ringhals PWR Units2016In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 182, no 2, p. 213-227Article in journal (Refereed)
    Abstract [en]

    Core-barrel motion (CBM) surveillance and diagnostics, based on the amplitude of the peaks of the normalized auto power spectral densities (APSDs) of the ex-core neutron detectors, have been performed and continuously developed in Sweden and were applied for monitoring of the three PWR units, Ringhals 2 to 4. From 2005, multiple measurements were taken during each fuel cycle, and these revealed a periodic behavior of the 8-Hz peak of the beam-mode motion: the amplitude increases within the cycle and returns to a lower value at the beginning of the next cycle. The work reported in this paper aims to clarify the physical reason for this behavior. A combination of a mode separation method in the time domain and a nonlinear curve fitting procedure of the frequency spectra revealed that two types of vibration phenomena contribute to the beam-mode peak. The lower frequency peak around 7 Hz in the ex-core detector APSDs corresponds to the CBM, whose amplitude does not change during the cycle. The higher frequency peak around 8 Hz arises from the individual vibrations of the fuel assemblies, and its amplitude increases monotonically during the cycle. This paper gives an account of the work that has been made to veri,b, the above hypothesis.

  • 11. Pillon, Sylvie
    et al.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Oxide and nitride TRU fuels: Lessons drawn from the CONFIRM and FUTURE projects of the 5th European Framework Program2006In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 153, no 3, p. 245-252Article in journal (Refereed)
    Abstract [en]

    The FUTURE and CONFIRM projects of the 5th European Framework Program address the issues of the design and fabrication of oxide and nitride fuels, respectively, for the transmutation in an accelerator-driven system (ADS). They started in December 2001 and September 2000, respectively. Advantages and drawbacks of transuranic oxides and nitrides in terms of performance and fabricability have been analyzed. Recommendations on the fuel design will be given and used for the next step of the 6th European Framework Program related to the design and the feasibility assessment of an industrial ADS prototype dedicated to transmutation.

  • 12.
    Seltborg, Per
    et al.
    KTH, School of Engineering Sciences (SCI), Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Proton source efficiency for heterogeneous distribution of actinides in the core of an accelerator-driven system2006In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 154, no 2, p. 202-214Article in journal (Refereed)
    Abstract [en]

    The distribution of actinides in the core of an accelerator-driven system loaded with plutonium, americium, and curium has been studied in order to optimize the proton source efficiency psi*. The optimization of psi* was performed by keeping some important characteristics of the system, e.g., the radial power profile and the reactivity of the core, constant. One of the basic assumptions of the study, that the magnitude of psi* is sensitive primarily to the composition of actinides in the inner part of the core, whereas only marginally to that in the outer part, has been confirmed. It has been shown that the odd-N nuclides (those nuclides with an even number of neutrons) in general and Am-241 and Cm-244 in particular have favorable properties with respect to improving psi* if they are placed in the innermost part of the core. The underlying reason for this phenomenon is that the energy spectrum of the source neutrons in the inner part of the core is harder than that of the average fission neutrons. Moreover, it has been shown that loading the inner part of the core with only curium increases psi* by similar to 7%. Plutonium, on the other hand, in particular high-quality plutonium consisting mainly of Pu-239 and Pu-241, was found to be a comparatively source inefficient element and is preferably located in the outer part of the core. The differences in psi* are due to combined effects from relative changes in the average fission and capture cross sections and in the average fission neutron yield.

  • 13.
    Seltborg, Per
    et al.
    KTH, Superseded Departments, Physics.
    Wallenius, Janne
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Tucek, Kamil
    KTH, Superseded Departments, Physics.
    Gudowski, Waclaw
    KTH, Superseded Departments, Physics.
    Definition and application of proton source efficiency in accelerator driven systems2003In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 145, no 3, p. 390-399Article in journal (Refereed)
    Abstract [en]

    In order to study the beam power amplification of an accelerator-driven system (ADS), a new parameter, the proton source efficiency psi* is introduced. psi* represents the average importance of the external proton source, relative to the average importance of the eigenmode production, and is closely related to the neutron source efficiency rho*, which is frequently used in the ADS field. rho* is commonly used in the physics of subcritical systems driven by any external source (spallation source, (d,d), (d, t), Cf-252 spontaneous fissions, etc.). On the contrary, psi* has been defined in this paper exclusively for ADS studies where the system is driven by a spallation source. The main advantage with using psi* instead of rho* for ADS is that the way of defining the external source is unique and that it is proportional to the core power divided by the proton beam power, independent of the neutron source distribution.

    Numerical simulations have been performed with the Monte Carlo code MCNPX in order to study psi* as a function of different design parameters. It was found that, in order to maximize psi* and therefore minimize the proton current needs, a target radius as small as possible should be chosen. For target radii smaller than similar to30 cm, lead-bismuth is a better choice of coolant material than sodium, regarding the proton source efficiency, while for larger target radii the two materials are equally good. The optimal axial proton beam impact was found to be located similar to 20 cm above the core center. Varying the proton energy, psi*/E-p was found to have a maximum for proton energies between 1200 and 1400 MeV Increasing the americium content in the fuel decreases psi* considerably, in particular when the target radius is large.

  • 14.
    Talamo, Alberto
    KTH, School of Engineering Sciences (SCI), Physics.
    Analytical calculation of the average Dancoff factor for prismatic high-temperature reactors2007In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 156, no 3, p. 343-356Article in journal (Refereed)
    Abstract [en]

    In the present studies we performed the analytical calculation of the average Dancoff factor for prismatic high-temperature reactors; in this type of core, the fuel elements consist of small fuel grains (TRISO particles) randomly dispersed in a moderator (graphite) matrix and confined to a cylindrical volume (fuel pin). By definition, the Dancoff factor is the probability that a neutron leaving a fuel kernel hits uncollided another fuel kernel in the same fuel pin, which represents the intrapin contribution, or in another pin, which represents the interpin contribution. Similar studies have already been performed for pebble bed high-temperature reactors, where spheres (fuel pebbles) play the role of the cylinders; consequently, we retained the physical model describing an infinite lattice of unit cells, each containing a pair of concentric spheres, where the inner sphere is filled with a mixture of fuel grains and moderator and the outer one is filled with pure moderator, and we derived the mathematical model for the case of concentric cylinders. The physical model is grounded on the chord theory and the concept of a pseudo cross section; the latter takes into account, when the medium consists of moderator and small fuel grains, the probability, per unit path length, that a neutron either collides with a moderator nucleus or hits a fuel surface. The above method possesses a general validity, and it is suitable for the treatment of spheres (fuel pebbles), cylinders (fuel pins), or cuboids (fuel prisms) filled by moderator and small fuel grains. The predictions of the analytical method well match the results of the MCNP code; nevertheless, since in the case of prismatic cores the mathematical model involves the calculation of complicated double integrals, the CPU time required by the two different methods becomes comparable.

  • 15.
    Talamo, Alberto
    et al.
    KTH, School of Engineering Sciences (SCI), Physics.
    Gudowski, Waclaw
    KTH, School of Engineering Sciences (SCI), Physics.
    Incineration of light water reactor waste in high-temperature gas reactors: Axial fuel management and efficiency of americium and curium transmutation2007In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 156, no 2, p. 244-266Article in journal (Refereed)
    Abstract [en]

    In the present study we investigate the influence of the fuel axial shuffling and the operational control rod maneuvering on the performances of the one-pass (no reprocessing) deep-burn incineration of light water reactor waste in the gas turbine-modular helium reactor. After an irradiation period, the fuel axial shuffling schedule has to take into account the fuel depletion profile generated by the adjustments of the position of the operational control rods, because the insertion of the rods strongly alters the neutron flux shape. We aimed at implementing a numerical simulation as close as possible to a real scenario and therefore took advantage of the powerful geometrical modeling capability of the MCB code to describe the reactor in a detailed three-dimensional geometry model in which we simulated over 120 different burnable materials, each of them undergoing a different neutron flux intensity. We adjusted the position of the control rods every 90 effective full-power days of irradiation to maintain the core as close as possible to the critical condition; thereafter, we recalculated the neutron flux and cross sections by a new MCNP/ MCB run. At the present time, this sophisticated approach can be realized only by a computer cluster of ten 64-bit processors working in parallel mode. The fuel axial shuffling adds from 3 to 5% to the transmutation rates of 239Pu, plutonium, and all actinides, which range from 80 to 86, 50 to 53, and 46 to 48%, respectively; the present results are 5 to 14% less compared to the case of a two-pass (reprocessing) deep burn. The efficiency of transmuting minor actinides has been estimated by comparing the long-term radio-toxicity of the fresh and irradiated americium and curium fuel; this comparison revealed that it is not worthwhile to transmute americium and curium in the current design of the gas turbine-modular helium reactor by a one-pass deep burn.

  • 16.
    Talamo, Alberto
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Gudowski, Wacław
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    A deep burn fuel management strategy for the incineration of military plutonium in the gas turbine-modular helium reactor modeled in a detailed three-dimensional geometry by the Monte Carlo continuous energy burnup code2006In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 153, no 2, p. 172-183Article in journal (Refereed)
    Abstract [en]

    In the future development of nuclear energy, the graphite-moderated helium-cooled reactors may play an important role because of their valuable technical advantages: passive safety, low cost, flexibility in the choice of fuel, high conversion energy efficiency, high burnup, more resistant fuel cladding, and low power density. General Atomics possesses a long experience with this type of reactor, and it has recently developed the gas turbine-modular helium reactor (GT-MHR), a design where the nuclear power plant is structured into four reactor modules of 600 MW(thermal). Amid its benefits, the GT-MHR offers a rather large flexibility in the choice of fuel type; Th, U, and Pu may be used in the manufacture of fuel with some degrees of freedom. As a consequence, the fuel management may be designed for different objectives aside from energy production, e.g., the reduction of actinide waste production through a fuel based on thorium. In our previous studies we analyzed the behavior of the GT-MHR with a plutonium fuel based on light water reactor (LWR) waste; in the present study we focused on the incineration of military Pu. This choice of fuel requires a detailed numerical modeling of the reactor since a high value of keff at the beginning of the reactor operation requires the modeling both of control rods and of burnable poison; by contrast, when the GT-MHR is fueled with LWR waste, at the equilibrium of the fuel composition, the reactivity swing is small.

  • 17.
    Wallenius, Janne
    et al.
    KTH, School of Engineering Sciences (SCI), Physics, Reactor Physics.
    Tucek, K.
    Carlsson, J.
    Gudowski, Waclaw
    KTH, Superseded Departments, Physics.
    Application of burnable absorbers in an accelerator-driven system2001In: Nuclear science and engineering, ISSN 0029-5639, E-ISSN 1943-748X, Vol. 137, no 1, p. 96-106Article in journal (Refereed)
    Abstract [en]

    The application of burnable absorbers (BAs) to minimize power peaking, reactivity loss, and capture-to-fission probabilities in an accelerator-driven waste transmutation system has been investigated. Boron-IO-enriched B4C absorber rods were introduced into a lead-bismuth-cooled core fueled with transuranic (TRU) discharges from light water reactors to achieve the smallest possible power peakings at beginning-of-life (BOL) subcriticality level of 0.97 Detailed Monte Carlo simulations show that a radial power peaking equal to 1.2 at BOL is attainable using a four-zone differentiation in BA content. Using a newly written Monte Carlo burnup code, reactivity losses were calculated to be 640 pcm per percent TRU burnup for unrecycled TRU discharges. Comparing to corresponding values in BA-free cores, BA introduction diminishes reactivity losses in TRU-fueled subcritical cores by similar to 20%. Radial power peaking after 300 days of operation at 1200-MW thermal power was0.92, which appears to be acceptable, with respect to limitations in cladding and fuel temperatures. In addition, the else of BAs yields significantly higher fission-to-capture probabilities in even-neutron-number nuclides. Fission-to-absorption probability ratio for Am-241 equal to 0.33 was achieved in the configuration studied. Hence, production of the strong alpha-emitter Cm-242 is reduced, leading to smaller fuel-swelling rates and pin pressurization. Disadvantages following BA introduction such as increase of void worth and decrease of Doppler feedback in conjunction with small values of beta (eff), need to be addressed by derailed studies of subcritical core dynamics.

1 - 17 of 17
CiteExportLink to result list
Permanent link
Cite
Citation style
  • apa
  • harvard1
  • ieee
  • modern-language-association-8th-edition
  • vancouver
  • Other style
More styles
Language
  • de-DE
  • en-GB
  • en-US
  • fi-FI
  • nn-NO
  • nn-NB
  • sv-SE
  • Other locale
More languages
Output format
  • html
  • text
  • asciidoc
  • rtf