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Sökning: L773:0029 5639 OR L773:1943 748X

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1.
  • Andersson, Mikael, 1986, et al. (författare)
  • Control Rod Calculation in Axially-Heterogeneous Fast Reactors. Part I: Influence of the Absorber Environment
  • 2017
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 185:2, s. 263-276
  • Tidskriftsartikel (refereegranskat)abstract
    • In axially heterogeneous fast reactor concepts, such as the ASTRID CFVcore, the accurate neutronic prediction of control rods is a challenge. In suchcores, the performance of the classical 2D equivalence procedure, used forcontrol rod homogenization in homogeneous fast reactors, is questionable.In this work (Part I of II), a number of axially heterogeneous environ-ments, representative of a CFV-type core are investigated using 2D (X-Z )models, with the objective to distinguish regions where the classical equiva-lence procedure is valid from those where it is not.It is found that the environments that affect the control rod absorber themost, and are likely to invalidate the procedure, are the internal control rodinterfaces, such as the absorber/follower interface and the interface betweenzones of different boron enrichments. The range of the main spectral impactcould be seen within 0-10 cm from the material interfaces studied.In a companion Paper (Part II), a full core investigation is performed,which builds upon the results of this paper.
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2.
  • Andersson, Mikael, 1986, et al. (författare)
  • Control rod calculation in axially-heterogeneous fast reactors. Part II: Impact of 3D homogenization on core parameters
  • 2017
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 185:2, s. 277-293
  • Tidskriftsartikel (refereegranskat)abstract
    • Advanced sodium-cooled fast reactors with improved safety features, such asthe French ASTRID CFV-core concept, is characterized by an axial heteroge-neous core, which will present a challenge on the homogenization proceduresused today, taking into account all the different axial material transitions.Reliable modeling of the control rod and accurate prediction of the controlrod worth are essential to determine the shutdown margins and to ensuresafe operation.In this work (Part II of II), two different homogenization schemes are com-pared. One is based on the traditional reactivity-equivalence procedure in 2D,and the other a newly implemented 3D version of the reactivity-equivalenceprocedure, with approximations based on the results in a companion pa-pers (Part I). The deterministic results are compared with a Monte Carloreference.Both of the cross section sets, from the two homogenization schemes, yielded results within the requested 5% error margin in reactivity. Thelargest discrepancy was found for the classical procedure for the case with aslightly inserted control rod (normal operating conditions).Both sets of cross sections yielded similar power profiles in the fuel sub-assembly neighboring the control rod within the 2 Monte Carlo standarddeviation. Neither of the cross section sets were able to predict the largegradients in capture rates close to the internal control rod interfaces.The study showed that the traditional 2D reactivity-equivalence proce-dure produces homogenized cross sections which yield reliable results in aCFV-type core. One exception from this was found for slightly inserted con-trol rods, where the effect of the follower/absorber interface could not be fullycaptured by the 2D scheme, and for such cases, 3D modeling is recommended.
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3.
  • Andersson, Mikael, 1986, et al. (författare)
  • Influence of Local Spectral Variations on Control-Rod Homogenization in Fast Reactor Environments
  • 2015
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 181:2, s. 204-215
  • Tidskriftsartikel (refereegranskat)abstract
    • Advanced fast reactor concepts, such as the CFV core (French acronym of "Coeur a Foible effet de Vide Sodium," meaning "low sodium void effect core"), are characterized by a heterogeneous axial core arrangement, with an inner fertile zone and a sodium plenum above the fuel. Such concepts represent a strong challenge for accurate predictions of the control-rod antireactivity effects, and the surrounding local fuel pin power. Classical equivalence procedures, which were developed for axially homogeneous cores, are put to the test when applied to such axially heterogeneous cores. In this work, we investigate the influence of variations in the local neutron spectra, for different control-rod environments, with the objective of understanding the impact of spectral variations in control-rod homogenization. This was conducted by considering a simple one-dimensional model of the equivalence procedure in which a transition zone between the fuel and control rod was introduced to represent different control-rod environments. Two types of situations were studied, one corresponding to softened neutron spectrum environments, for which the impact in the homogenized control-rod cross section was found to be smaller than 5%. The second situation was with wide elastic scattering resonances in the control-rod environment, which could locally lead to differences of up to 15% in the resulting equivalent cross sections. The reactivity effect of these changes was calculated to be less than 2%. In some cases, the numerical stability of the equivalence procedure was adversely affected, mainly in high-energy groups, due to the softening of the neutron spectra.
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4.
  • Anglart, Henryk, et al. (författare)
  • Fluid mechanics of Taylor bubbles and slug flows in vertical channels
  • 2002
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 140:2, s. 165-171
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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5.
  • Berglöf, Carl, et al. (författare)
  • Spatial and Source Multiplication Effects on the Area Ratio Reactivity Determination Method in a Strongly Heterogeneous Subcritical System
  • 2010
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 166:2, s. 134-144
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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6.
  • Chernikova, Dina, 1982, et al. (författare)
  • The Inclusion of Photofission, Photonuclear, (n, xn), (n, n ' x gamma), and (n, x gamma) Reactions in the Neutron-Gamma Feynman-Alpha Variance-to-Mean Formalism
  • 2017
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 185:1, s. 206-216
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper sets up a formalism that is sufficiently general to describe the effects of photofission, photonuclear, (n, xn), (n, n?x?), and (n, x?) reactions on the neutron-gamma Feynman-alpha variance-to-mean ratios. Such a formalism is obtained using the Chapman-Kolmogorov (master) forward equation for the above-mentioned set of nuclear reactions. Thereafter, the issue of estimating reaction intensities for gammas in the master equation is highlighted by the paper. As an example, a quantitative evaluation of reaction intensities is given for a case when (n, ?), photonuclear, and (n, 2n) reactions are relevant for the system. However, an evaluation of the influence of these types of reactions to the values of the Feynman variance-to-mean ratios is not within the scope of this paper. Overall, the results obtained in this paper are intended to give an extended systematic framework for the study of the neutron- and gamma-based nondestructive assay problems in nuclear reactor applications and materials control.
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7.
  • Cvetkovic, Vladimir, et al. (författare)
  • Comparative measures of radionuclide containment in the crystalline geosphere
  • 2002
  • Ingår i: Nuclear science and engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 142:3, s. 292-304
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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8.
  • Degweker, S. B., et al. (författare)
  • Stochastic Invariant Imbedding Theory for a Distributed Internal Source
  • 2011
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 168:3, s. 248-264
  • Tidskriftsartikel (refereegranskat)abstract
    • Invariant imbedding theory is an alternative formulation of particle transport theory. Until very recently, this theory was used only for deterministic calculations, i.e., for calculations of the first moment of the particle distribution. In a previous paper we set up a probability balance equation in the invariant imbedding approach. An equation was also obtained for the probability generating functional (pgfl) of reflected particles from which equations for the first- and second-order densities were derived. The approach was illustrated by a simple forward-backward scattering model with and without incorporating energy dependence to describe sputtering due to an external source of energetic particles on a medium. In this paper we extend these results to the case of a distributed internal source of particles. Among the possible applications, we discuss the problem of internal sputtering. We derive equations for the pgfl and the first- and second-order densities and show their connection with the external source problem. We treat the finite slab problem in addition to the semi-infinite slab geometry considered in our previous paper.
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9.
  • Demaziere, Christophe, 1973, et al. (författare)
  • Development of a method for measuring the MTC by noise analysis and its experimental verification in Ringhals-2
  • 2004
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 148:1, s. 1-29
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper deals with the estimation of the moderator temperature coefficient of reactivity (MTC) by noise analysis. The current noise analysis-based MTC measurement, relying on the cross correlation between the neutron noise measured by a single in-core neutron detector and the local temperature noise given by a single core-exit thermocouple located at the top of the same fuel assembly, or of a neighboring fuel assembly, is not accurate. The MTC is systematically underestimated by a factor of 2 to 5 compared to its design-predicted value. A theoretical study shows that, in case of nonhomogeneous moderator temperature noise, the core-averaged moderator temperature noise should be used for the MTC estimation. The new estimation method can reach up to 3% accuracy as compared with the results of core calculations for the Swedish Ringhals-2 pressurized water reactor (PWR). We show via noise measurements performed at the Ringhals-2 PWR that the moderator temperature noise is actually radially strongly heterogeneous and loosely coupled. The new MTC noise estimator is demonstrated to provide an accurate MTC evaluation, with the core-averaged moderator temperature noise estimated via the use of many radial in-core gamma-thermometers. More important, different forms of weighting functions are suggested to calculate the core-averaged moderator temperature noise. This new MTC noise estimator, which is nonintrusive and free of calibration, can therefore be applied to monitor the MTC throughout the cycle.
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10.
  • Demaziere, Christophe, 1973, et al. (författare)
  • Multi-fractal analysis of chaotic flashing-induced instabilities in boiling channels in the natural circulation CIRCUS facility
  • 2008
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 158:2, s. 164-193
  • Tidskriftsartikel (refereegranskat)abstract
    • In this paper, two-phase flow oscillations at the natural circulation CIRCUS test facility are investigated in a two-riser configuration. These oscillations are driven by flashing (and to some extent by geysering). For a given range of operating conditions of the facility, the oscillations exhibit an erratic behaviour. This study demonstrates that this behaviour can be attributed to deterministic chaos. This is proven by performing a Continuous Wavelet Transform of the measured time series. Any hidden self-similarity in the measurement is seen in the corresponding scale-space plane. The novelty of the present investigation lies with the multi-fractal approach used for characterizing the chaos. Both non-linear time series analysis and wavelet-based analysis methods show that the dynamics of the flow oscillations has a multi-fractal structure. For the former, both Higuchi’s method and Detrended Fluctuation Analysis were used, whereas for the latter, the Wavelet-Transform Modulus-Maxima method was used. The strange attractor corresponding to the dynamics of the system can thus be described as a set of interwoven mono-fractal objects. The global singular properties of the measured time series is then fully characterized by a spectrum of singularities f(alpha), which is the Hausdorff dimension of the set of points where the multi-fractal object has singularities of strength (or Hölder exponents of) alpha. Whereas Higuchi’s method and Detrended Fluctuation Analysis allow easily determining whether the deterministic chaos has a mono- or multi-fractal hierarchy, the Wavelet-Transform Modulus-Maxima has the advantage of giving a quantitative estimation of the fractal spectrum. The time-modelling of such a behaviour of the facility is therefore difficult since there is sensitive dependence on initial conditions. From a regulatory point-of-view, such a behaviour of natural circulation systems in a multiple riser configuration has thus to be avoided.
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11.
  • Dufek, Jan, et al. (författare)
  • Numerical Stability of Existing Monte Carlo Burnup Codes in Cycle Calculations of Critical Reactors
  • 2009
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 162:3, s. 307-311
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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12.
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13.
  • Dufek, Jan, et al. (författare)
  • Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors
  • 2006
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 152, s. 274-283
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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14.
  • Eriksson, Marcus, et al. (författare)
  • On the Performance of Point Kinetics for the Analysis Accelerator-driven Systems
  • 2005
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 149:3, s. 298-311
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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15.
  • Helgesson, Petter, 1986-, et al. (författare)
  • UO-2 Versus MOX: Propagated Nuclear Data Uncertainty for k-eff, with Burnup
  • 2014
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 177:3, s. 321-336
  • Tidskriftsartikel (refereegranskat)abstract
    • Precise assessment of propagated nuclear data uncertainties in integral reactor quantities is necessary for the development of new reactors as well as for modified use, e.g. when replacing UO-2 fuel by MOX fuel in conventional thermal reactors.This paper compares UO-2 fuel to two types of MOX fuel with respect to propagated nuclear data uncertainty, primarily in k-eff, by applying the Fast Total Monte Carlo method (Fast TMC) to a typical PWR pin cell model in Serpent, including burnup. An extensive amount of nuclear data is taken into account, including transport and activation data for 105 isotopes, fission yields for 13 actinides and thermal scattering data for H in H2O.There is indeed a significant difference in propagated nuclear data uncertainty in k-eff; at 0 burnup the uncertainty is 0.6 % for UO-2 and about 1 % for the MOX fuels. The difference decreases with burnup. Uncertainties in fissile fuel isotopes and thermal scattering are the most important for the difference and the reasons for this are understood and explained.This work thus suggests that there can be an important difference between UO-2 and MOX for the determination of uncertainty margins. However, the effects of the simplified model are difficult to overview; uncertainties should be propagated in more complicated models of any considered system. Fast TMC however allows for this without adding much computational time.
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16.
  • Ignatyuk, A. V., et al. (författare)
  • Neutron and proton cross-section evaluations for Th-232 up to 150 MeV
  • 2002
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 142:2, s. 177-194
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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17.
  • Ignatyuk, A. V., et al. (författare)
  • Neutron cross-section evaluations for U-238 up to 150 MeV
  • 2000
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 136:3, s. 340-356
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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18.
  • Insulander Björk, Klara, 1982, et al. (författare)
  • Scoping Studies of Dopants for Stabilization of Uranium Nitride Fuel
  • 2019
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 193:11, s. 1255-1264
  • Tidskriftsartikel (refereegranskat)abstract
    • Uranium nitride (UN) is considered as nuclear reactor fuel because of, among other reasons, its high uranium density and its high thermal conductivity. Its main drawback is that it relatively easily dissolves in hot water, which is particularly problematic when it is used in water-cooled reactors. One possible remedy to this is to add some corrosion inhibitor as dopant to the UN matrix. A number of dopants have been identified that have the potential to inhibit the dissolution process, and their respective merits have been investigated both by neutronic simulations and dissolution experiments. It is concluded that chromium is the most promising candidate.
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19.
  • Jansson, Peter, 1971-, et al. (författare)
  • A Device for Nondestructive Experimental Determination of the Power Distribution in a Nuclear Fuel Assembly
  • 2006
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 152:1, s. 76-86
  • Tidskriftsartikel (refereegranskat)abstract
    • There is a general interest in experimentally determining the power distribution in nuclear fuel. The prevalent method is to measure the distribution of the fission product 140Ba, which represents the power distribution over the last few weeks. In order to obtain the rod-by-rod power distribution, the fuel assemblies have to be dismantled.In this paper, a device for experimental nondestructive determination of the thermal rod-by-rod power distribution in boiling water reactor and pressurized water reactor fuel assemblies is described. It is based on measurements of the 1.6-MeV gamma radiation from the decay of 140Ba/La and utilizes a tomographic method to reconstruct the rod-by-rod source distribution. No dismantling of the fuel assembly is required.The device is designed to measure an axial node in 20 min with a precision of >2% (1). It is primarily planned to be used for validation of production codes for core simulation but may also be used for safeguards purposes.
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20.
  • Jansson, Peter, et al. (författare)
  • Blind Benchmark Exercise for Spent Nuclear Fuel Decay Heat
  • 2022
  • Ingår i: Nuclear science and engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 196:9, s. 1125-1145
  • Tidskriftsartikel (refereegranskat)abstract
    • The decay heat rate of five spent nuclear fuel assemblies of the pressurized water reactor type were measured by calorimetry at the interim storage for spent nuclear fuel in Sweden. Calculations of the decay heat rate of the five assemblies were performed by 20 organizations using different codes and nuclear data libraries resulting in 31 results for each assembly, spanning most of the current state-of-the-art practice. The calculations were based on a selected subset of information, such as reactor operating history and fuel assembly properties. The relative difference between the measured and average calculated decay heat rate ranged from 0.6% to 3.3% for the five assemblies. The standard deviation of these relative differences ranged from 1.9% to 2.4%.
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21.
  • Kozlowski, Tomasz, et al. (författare)
  • Cell Homogenization Method for Pin-by-Pin Neutron Transport Calculations
  • 2011
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 169:1, s. 1-18
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
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22.
  • Loberg, John, et al. (författare)
  • Neutron Detection-Based Void Monitoring in Boiling Water Reactors
  • 2010
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 164:1, s. 69-79
  • Tidskriftsartikel (refereegranskat)abstract
    • The ratio between the thermal- and fast-neutron fluxes in a boiling water reactor depends on the void fraction. The density of the steam-water mixture present in the core determines the efficiency of the moderation of fast neutrons born in fission; therefore, the void fraction could be determined by means of a simultaneous measurement of the thermal- and fast-neutron fluxes. Such measurement could also be used to investigate channel bow of the nuclear fuel bundles surrounding the detector because of sensitivity of the thermal flux to geometry changes. Calculations have been performed with both lattice and nodal codes to study the behavior of the void fraction correlation to the ratio of the thermal- and fast-neutron fluxes. The results prove the correlation to be nearly linear and robust. The rate of change of the correlation is insensitive to standard reactor operating parameters such as control rods and burnable absorbers; the sensitivity of the ratio to void fraction changes primarily depends on the geometry of the fuel bundles. A linear prediction model was used to represent the nodal code results. The absolute void fraction at over 792 positions in the core could be predicted with an absolute uncertainty of +/- 1.5%.
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23.
  • Loberg, John, 1980-, et al. (författare)
  • Simulations and Models of Neutron Fluxes in BWRs Intended for Depletion Calculations of Withdrawn Control Rods
  • 2011
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 177:3, s. 221-229
  • Tidskriftsartikel (refereegranskat)abstract
    • Models of the neutron flux shape in a withdrawn control rod in a boiling water reactor (BWR) bottom reflector have been constructed from simulations with the Monte Carlo code MCNP. These neutron flux models are intended for determining absorber depletion and fast fluence accumulation for withdrawn control rods with nodal codes. So-called G-factors are created for coupling the neutron flux models to a conventional nodal code via the core bottom neutron flux. The neutron flux models and G-factors are created for three different neutron energies, and their dependence on various parameters such as blanket enrichments, Hf and B(4)C control rod absorber, and depletion and reflector geometry is investigated. The neutron flux models and G-factors are found to be very insensitive; the neutron flux models predict the simulated neutron flux in the withdrawn control rod from MCNP over a variety of reflector configurations with an error < 3.0%. This implies that the neutron flux models constructed in this paper are generally applicable for BWR reflectors and control rods not fundamentally deviating from the designs investigated in this paper.
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24.
  • Lundqvist Saleh, Tobias, et al. (författare)
  • Recent Progress in the Design of a Tomographic Device for Measurements of the Three-Dimensional Pin-Power Distribution in Irradiated Nuclear Fuel Assemblies
  • 2010
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 165:2, s. 232-239
  • Tidskriftsartikel (refereegranskat)abstract
    • A tomographic technique for determination of the thermal power distribution in nuclear fuel assemblies is under development. The purpose is to provide an experimental validation tool for core simulation codes. Such codes are essential for the operation of nuclear power reactors, and validation is important in the process of improving and developing the codes as well as the fuel. The tomographic method is nonintrusive and offers large amounts of data within a normal revision shutdown. In earlier experimental investigations using a test platform, the method proved useful, demonstrating results of satisfying quality. However, the measuring setup also revealed nonfeasible properties related to transport, decontamination, and background radiation shielding. In this paper, the design of a new measuring device is presented. It is based on experiences from the test platform, but its size and weight make it advantageous regarding transports and decontamination. Moreover, the design inherently allows for more efficient background shielding. The latter has been investigated in a detailed study using the MCNP simulation code. The results confirm the high levels of background radiation observed in the test platform. It is also concluded that the shielding properties in the new design are sufficient.
  •  
25.
  • Nyqvist, Robert, 1983, et al. (författare)
  • Spreading of collimated particle beams within a generalized Fokker-Planck diffusion equation
  • 2009
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 163:1, s. 85-90
  • Tidskriftsartikel (refereegranskat)abstract
    • Recently, an expansion of the Boltzmann scattering operator describing the angular spreading of particle beams was given that included the effects of large angle scattering processes, thus generalizing the classical Fokker-Planck equation, valid in the limit of small angle scattering. The present work aims at making an analytical comparison between predictions based on the classical Fokker-Planck equation and those based on a generalized one, which includes a first-order correction term in the expansion of the Boltzmann scattering operator. The analysis is carried out for thin slabs where backscattering effects can be neglected and makes use of a moment approach, which leads to an infinite system of recursively coupled ordinary differential equations. The system is truncated in a consistent manner, and the effects of large angle scattering on the evolution of the moments are determined in explicit analytical form. An approximate similarity solution of the generalized Fokker-Planck equation is also found, and the results of both approaches provide a clear picture of the increased diffusive beam spreading due to large angle scattering. A comparison with previously published Monte Carlo simulation results shows good agreement.
  •  
26.
  • Osifo, Otasowie, et al. (författare)
  • Verification and determination of the decay heat in spent PWR fuel by means of gamma scanning
  • 2008
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 160:1, s. 129-143
  • Tidskriftsartikel (refereegranskat)abstract
    • Decay heat is an important design parameter at the future Swedish spent nuclear fuel repository. It will be calculated for each fuel assembly using dedicated depletion codes, based on the operator-declared irradiation history. However, experimental verification of the calculated decay heat is also anticipated. Such verification may, be obtained by, gamma scanning using the established correlation between the decay heat and the emitted gamma-ray intensity from Cs-137. In this procedure, the correctness of the operator-declared fuel parameters can be verified. Recent achievements of the gamma-scanning technique include the development of a dedicated spectroscopic data-acquisition system and the use of an advanced calorimeter for calibration. Using this system, the operator-declared burnup and cooling time of 31 pressurized water reactor fuel assemblies was verified experimentally, to within 2.2% (1 sigma) and 1.9% (1 sigma), respectively. The measured decay heat agreed with calorimetric data within 2.3% (1 sigma). whereby the calculated decay, heat was verified within 2.3% (1 sigma). The measuring time per fuel assembly was similar to 15 min. In case reliable operator-declared data are not available, the gamma-scanning technique also provides a means to independently measure the decay, heat. The results obtained in this procedure agreed with calorimetric data within 2.7% (1 sigma).
  •  
27.
  • Pal, L., et al. (författare)
  • Stochastic Theory of the Fission Chamber Current Generated by Non-Poissonian Neutrons
  • 2016
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 184:4, s. 537-550
  • Tidskriftsartikel (refereegranskat)abstract
    • The Campbell theorem, relating the variance of the current of a fission chamber (a "filtered Poisson process") to the intensity of the detection events and to the detector pulse shape, becomes invalid when the neutrons generating the fission chamber current are not independent. Recently, a formalism was developed by the present authors, by which the variance of the detector current can be calculated for detecting neutrons in a subcritical multiplying system, where the detection events are obviously not independent. In the present paper, the previous formalism, which only accounted for prompt neutrons, is generalized to account also for delayed neutrons. A rigorous probabilistic analysis of the detector current was performed by using the same simple, but realistic detector model as in the previous work. The results of the present analysis made it possible to determine the bias of the traditional Campbelling techniques both qualitatively and quantitatively. The results show that the variance still remains proportional to the detection intensity, and is thus suitable for the monitoring of the mean flux, but the calibration factor between the variance and the detection intensity is an involved function of the detector pulse shape and the subcritical reactivity of the system, which diverges for critical systems.
  •  
28.
  • Pal, L., et al. (författare)
  • The fast fission factor revisited
  • 2009
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 161:1, s. 111 - 118
  • Tidskriftsartikel (refereegranskat)abstract
    • The concept and calculation techniques of multiplicities innuclear safeguards are applied to the calculation of the traditional fast fission factor of reactor physics. The concept is the assumption that the original source neutrons from spontaneous or induced fission, and the further neutrons given rise through fast fission in the sample before leakage, are considered as being generated simultaneously with the source neutrons. The numberdistribution of the neutrons arising from such a "superfission" process will be different from that of the nuclear fission process. Concerning the mathematical treatment, in safeguards works the master equation approach is used to calculate the moments of such a distribution. Hence, to follow suite, a derivation of the fast fission factor is given here by a backward master equation. This method has the advantages that the derivation of the fast fission factor becomes more transparent than the traditional one, and that it allows also to determinehigher order moments, notably the variance, of the total number of neutrons generated, i.e. when account is taken also of the contribution of fast fission to these moments. The results show that the relative standard deviation increases fast with the increase of the non-leakage probability of neutrons, and hence with the increase of the fast fission factor itself. Also the Diven factor of the superfission process (neutrons from fast fissions included) is significantly larger than that of thermal fission. We argue that the traditional model, in which the Feynman- and Rossi-alpha models are derived, does not account correctly for the extra branching represented by the fast fissionprocess. Hence the Diven factor traditionally used in thoseformulae should be used in a modified form. We show how the effect of fast fission needs to be included into the model to obtain the correct formula, and give explicit expressions. Some quantitative examples are given for illustration.
  •  
29.
  • Pal, L., et al. (författare)
  • Theory of neutron noise in a temporally fluctuating multiplying medium
  • 2007
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 155:3, s. 425-440
  • Tidskriftsartikel (refereegranskat)abstract
    • Neutron fluctuations in a constant multiplying medium (zero power noise) and those in a fluctuating medium (power reactor noise) have been traditionally considered as two separate disciplines that exist in two opposing limiting areas of operation (low and high power, respectively). They have also been treated by different mathematical methods, i.e., master equations and Langevin equation, respectively. In this paper we develop a theory of neutron fluctuations in a medium randomly varying in time, based on a forward-type master equation approach. This method accounts for both the zero power and the power reactor noise simultaneously. Factorial moments and related quantities (variance, power spectrum, etc.) of the number of the neutrons are calculated in subcritical systems with a stationary external source. It is shown that the pure zero power and power reactor noise results can be reconstructed in the cases of vanishing system fluctuations and high power, respectively, the latter being a nontrivial result. Further, it is shown that the effect of system fluctuations on the zero power noise is retained even in the limit of vanishing neutron number (reactor power). The results have thus even practical significance for low-power systems with fluctuating properties. The results also have a bearing on other types of branching processes such as evolution of biological systems, germ colonies, epidemics, etc., which take place in a time-varying environment.
  •  
30.
  • Pazsit, Imre, 1948, et al. (författare)
  • Developments in Core-Barrel Motion Monitoring and Applications to the Ringhals PWR Units
  • 2016
  • Ingår i: Nuclear science and engineering. - : American Nuclear Society. - 0029-5639 .- 1943-748X. ; 182:2, s. 213-227
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
31.
  • Pazsit, Imre, 1948, et al. (författare)
  • Reactor Kinetics, Dynamic Response, and Neutron Noise in Molten Salt Reactors
  • 2011
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 167, s. 61-76
  • Tidskriftsartikel (refereegranskat)abstract
    • The dynamic space- and frequency dependent response of a Molten SaltReactor (MSR) to stationary perturbations is investigated in asimple analytical model. The Green's function of the system isinvestigated in the general case of arbitrary fuel recirculationvelocity and in the limiting case of infinite fuel velocity whichpermits closed form solutions both in the static and dynamic case.It is found that the amplitude of the induced noise is generallyhigher and the domain of the point kinetic behaviour valid up tohigher frequencies than in a corresponding traditional system. Thisis due to the differing behaviour of the delayed neutron precursorsas compared to the traditional case. The MSR equations are notself-adjoint and the adjoint equation and adjoint function have tobe constructed, which is also done here. Finally the space-dependentneutron noise, induced by propagating perturbations of theabsorption cross section is calculated. A number of interestingproperties that are relevant to full size MSRs are found andinterpreted. The results are consistent with those in traditionalsystems but the domains of various behaviour regimes (point kinetic,space dependent etc.) are shifted to higher frequencies or systemsizes.
  •  
32.
  • Pazsit, Imre, 1948, et al. (författare)
  • Space-Dependent Calculation of the Multiplicity Moments for Shells With the Inclusion of Scattering
  • 2023
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 197:8, s. 2030-2046
  • Tidskriftsartikel (refereegranskat)abstract
    • In recent work, we extended the methodology of multiplicity counting in nuclear safeguards by elaborating the one-speed stochastic transport theory of the calculation of the so-called multiplicity moments, i.e., the factorial moments of the number of neutrons emitted from a fissile item, following a source event from an internal neutron source [spontaneous fission and ((Formula presented.)) reactions]. Calculations were made for solid spheres and cylinders, with the source being homogeneously distributed within the item. Recent measurements of the Rocky Flats Shells during the Measurement of Uranium Subcritical and Critical (MUSIC) campaign conducted by Los Alamos National Laboratory and assisted by the University of Michigan inspired us to extend the model to spherical shell geometry with a point source in the middle of the central cavity. Comparison of the calculated results with the experimental ones indicated that accounting for fission as the only neutron reaction (the standard procedure in the point model, adapted also in our work so far) was not sufficient for reaching good agreement with measurements. The model was therefore extended to include elastic scattering into the one-speed formalism, whereas the effect of inelastic scattering was accounted for in an empirical way. After these extensions, good agreement was found between the calculated and the measured values. The paper describes the extension of the theory and provides concrete quantitative results.
  •  
33.
  • Pazsit, Imre, 1948, et al. (författare)
  • Transport calculations of the multiplicity moments for cylinders
  • 2022
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 196:3, s. 235-249
  • Tidskriftsartikel (refereegranskat)abstract
    • In a previous paper by Pázsit and Pál [“Multiplicity Theory Beyond the Point Model,” Ann. Nucl. Energy, Vol. 154 (2021)], a general transport theory calculation of the factorial moments of the number of neutrons emitted spontaneously from a sample was elaborated. In contrast to the original derivations by Hage and Cifarelli [“On the Factorial Moments of the Neutron Multiplicity Distribution of Fission Cascades,” Nucl. Instrum. Meth. Phys. Res. A, Vol. 236 (1985)] and Böhnel [“The Effect of Multiplication on the Quantitative Determination of Spontaneously Fissioning Isotopes by Neutron Correlation Analysis,” Nucl. Sci. Eng., Vol. 90 (1985)], also referred to as the point model, in the transport model the spatial and angular dependence of the internal fission chain is taken into account with a one-speed transport theory treatment. Quantitative results were given for a spherical item, and the bias of the point model regarding the estimation of the fission rate as compared to the more exact space-dependent model was estimated as a function of the size of the sphere and the α factor. In the present paper the formalism and the quantitative work are extended to the treatment of items with cylindrical shapes, which are more relevant in many practical applications. Results are presented for both square cylinders (D=H) and for tall (H/D>1) and flat (H/D<1) cylinders. This way the differences between the cylinder and the sphere on one hand and those between the various cylinder shapes on the other hand can be estimated. The results show that the bias depends on the geometry of the cylinder quite moderately, but similarly to the case of the sphere, the bias of the point model is quite significant for larger item sizes and α values, and it is nonconservative (underestimates the fissile mass) as well.
  •  
34.
  • Pillon, Sylvie, et al. (författare)
  • Oxide and nitride TRU fuels : Lessons drawn from the CONFIRM and FUTURE projects of the 5th European Framework Program
  • 2006
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 153:3, s. 245-252
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
35.
  • Pozzi, Sara A., et al. (författare)
  • The Statistics of the Number of Neutron Collisions prior to Absorption
  • 2006
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 153:1, s. 60 - 68
  • Tidskriftsartikel (refereegranskat)abstract
    • We propose a simple analytical model to describe the statistics of the number of collisions undergone by fast neutrons during slowing down until they are absorbed. We assume that the moderator is homogeneous and account for scattering and absorption, but do not consider thermalization. Although the problem cannot be solved in a compact form, a simple recurrent formula provides the solution in a very transparent way. The model can be readily evaluated numerically, and the results are in excellent agreement with the corresponding Monte Carlo simulations. Both the mean number and the variance of the number of collisions are calculated. The results are discussed and compared with the classical case of neutron slowing down to or past a given energy in a moderating medium without absorption.
  •  
36.
  • Pozzi, Sara, et al. (författare)
  • Neutron Slowing Down in a Detector with Absorption
  • 2006
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 154:3, s. 367 - 373
  • Tidskriftsartikel (refereegranskat)abstract
    • In a recent paper, we presented a simple analytical model to describe the statistics of the number of scattering collisions undergone by fast neutrons as they slow down until they are absorbed. In that study, we assumed that the moderator was infinite and homogeneous, and accounted for scattering and absorption by a single nuclear species. In the present paper, we extend that methodology to the more realistic case of neutron slowing down in a homogeneous mixture. The formulae are derived and evaluated numerically, and the results are found to be in very good agreement with corresponding Monte Carlo simulations. The average value of the number of collisions that a neutron undergoes before being captured is computed. The results for a capture-gated detector composed of hydrogen, carbon, and boron are discussed.
  •  
37.
  • Qvist, Staffan, 1986-, et al. (författare)
  • Design Space Analysis for Breed-and-Burn Reactor Cores
  • 2016
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 183:2, s. 197-212
  • Tidskriftsartikel (refereegranskat)abstract
    • For a reactor to establish a sustainable breed-and-burn (B&B) mode of operation, its fuel has to reach a minimum level of average burnup. The value of the minimum required average discharge burnup strongly depends on the core design details. Using the extended neutron balance method, it is possible to quantify the impact of major core design choices on the minimum required burnup in a B&B core. Relevant design variables include the fuel chemical form, nonactinide mass fraction of metallic fuel, feed-fuel fissile fraction, fuel rod pitch-to-diameter ratio (P/D), average neutron flux level, and fraction of neutron loss. Metallic fuels have been found to be the only viable fuel options for a realistic near-term B&B reactor. For the core designs we have studied, it was not possible to sustain B&B operation using oxide fuel that is not enriched, while nitride and carbide fuels may only work in highly ideal low-leakage systems at very high levels of discharge burnup and, hence, neutron irradiation damage. The minimum required burnup increases strongly with the total fraction of neutrons that is lost to leakage and reactivity control. The flux level has no effect on the neutron balance within the applicable range, and the average discharge burnup is also relatively insensitive to the fraction of fissile material in the feed fuel in the range from depleted uranium (0.2% U-235) to natural uranium (0.71% U-235). The minimum required burnup is most sensitive, in order of importance, to the fractional loss of neutrons, the Zr content in metallic fuel, and the fuel rod P/D. Changing the weight fraction of zirconium in metallic fuel by 1% (for example, from 10% to 9%) gives the same change in required discharge burnup as adjusting the P/D by 0.02 (for example, from 1.10 to 1.12).
  •  
38.
  • Rochman, Dimitri, et al. (författare)
  • Efficient use of Monte Carlo : Uncertainty Propagation
  • 2014
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 177:3, s. 337-349
  • Tidskriftsartikel (refereegranskat)abstract
    • A new and faster Total Monte Carlo method for the propagation of nuclear data uncertaintiesin Monte Carlo nuclear simulations is presented (the fast TMC method).It is addressing the main drawback of the original Total Monte Carlo method(TMC), namely the necessary large time multiplication factor compared to a singlecalculation. With this new method, Monte Carlo simulations can now be accompaniedwith uncertainty propagation (other than statistical), with small additionalcalculation time. The fast TMC method is presented and compared with the TMCand fast GRS methods for criticality and shielding benchmarks and burn-up calculations.Finally, to demonstrate the efficiency of the method, uncertainties on localdeposited power in 12.7 millions cells are calculated for a full size reactor core,
  •  
39.
  • Santoro, V., et al. (författare)
  • The HighNESS Project at the European Spallation Source : Current Status and Future Perspectives
  • 2024
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 198:1, s. 31-63
  • Tidskriftsartikel (refereegranskat)abstract
    • The European Spallation Source (ESS), presently under construction in Lund, Sweden, is a multidisciplinary international laboratory that, once completed at full specifications, will operate the world's most powerful pulsed neutron source. Supported by a 3 M Euro Research and Innovation Action within the European Union Horizon 2020 program, a design study (HighNESS) is now underway to develop a second neutron source located below the spallation target. Compared to the first source, which is located above the spallation target and designed for high cold and thermal brightness, the new source is being optimized to deliver higher intensity and a shift to longer wavelengths in the spectral regions of cold neutrons (CNs) (2 to 20 & Aring;), very cold neutrons (VCNs) (10 to 120 & Aring;), and ultracold neutrons (UCNs) (> 500 & Aring;). The second source consists of a large liquid deuterium moderator to deliver CNs and serve secondary VCN and UCN sources, for which different options are under study. These new sources will boost several areas of condensed matter research and will provide unique opportunities in fundamental physics. The HighNESS project is now entering its last year, and we are working toward the Conceptual Design Report of the ESS upgrade. In this paper, results obtained in the first 2 years, ongoing developments, and future perspectives are described.
  •  
40.
  • Seltborg, Per, et al. (författare)
  • Definition and application of proton source efficiency in accelerator driven systems
  • 2003
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 145:3, s. 390-399
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
41.
  • Seltborg, Per, et al. (författare)
  • Proton source efficiency for heterogeneous distribution of actinides in the core of an accelerator-driven system
  • 2006
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 154:2, s. 202-214
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
42.
  • Shin, Tony H., et al. (författare)
  • Neutron multiplicity counting moments for fissile mass estimation in scatter-based neutron detection systems
  • 2017
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 188:3, s. 246-269
  • Tidskriftsartikel (refereegranskat)abstract
    • Neutron multiplicity counting (NMC) techniques are widely used for nuclear materials accountability and international safeguards applications to quantitatively evaluate characteristic properties pertaining to fissile material. Mathematical models for NMC moments have been previously derived for systems that use capture-based detectors; however, these models are not applicable when scatter-based detectors are used because of "neutron cross talk." Neutron cross talk occurs when a single neutron scatters and deposits energy above threshold into multiple detectors causing spurious increase in multiplicity counts; this, in turn, has caused fissile mass to be overestimated when not treated. In this paper, we propose new mathematical models derived from point kinetics to correct for neutron cross-talk effects up to any arbitrary order N, where N denotes the maximum number of counts a single neutron can cause. The new models were used to estimate the fissile mass of plutonium metal and oxide samples with effective 240Pu mass ranging from 2.5 to 250 g. The adequacy of the models was confirmed using simulations of a conceptual scatter-based neutron multiplicity counter (e.g., organic scintillators) using MCNPX v2.7e with the PoliMi fission event generating extension. The fissile mass estimates with no correction for neutron cross-talk events yielded an average relative deviation from the true 240Pueff mass of 55.94% and 84.56% for metal and oxide samples, respectively. When neutron cross-talk events of order N = 2 are included in the model, the fissile mass estimates yielded an average relative deviation of 11.89% for metal and 13.21% for oxide samples. Accounting for neutron cross-talk events of order N = 3 resulted in fissile mass estimates with an average relative deviation of 9.58% and 10.51% for metal and oxide samples, respectively. These mass estimates were compared to a reference case (i.e., no neutron crosstalk effects) that yielded an average relative deviation of 6.81% and 4.77% for metal and oxide samples, respectively. The discrepancy between the estimates from the proposed model and the reference case is attributed to the assumed value of N, which sets a finite upper bound on the order of cross-talk events the model treats (i.e., the model for N = 3 assumes that a neutron will never cause more than three counts).
  •  
43.
  • Talamo, Alberto, et al. (författare)
  • 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 code
  • 2006
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 153:2, s. 172-183
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
44.
  • Talamo, Alberto (författare)
  • Analytical calculation of the average Dancoff factor for prismatic high-temperature reactors
  • 2007
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 156:3, s. 343-356
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
45.
  • Talamo, Alberto, et al. (författare)
  • Incineration of light water reactor waste in high-temperature gas reactors : Axial fuel management and efficiency of americium and curium transmutation
  • 2007
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 156:2, s. 244-266
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
46.
  • Tran, Hoai Nam, 1981, et al. (författare)
  • Neutron noise calculations in hexagonal geometry and comparison with analytical solutions
  • 2013
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 175:3, s. 340-351
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper presents the development of a neutronic and kinetic solver for neutron noise calculations in hexagonal geometries. The tool is developed based on diffusion theory with multienergy groups and several groups of delayed neutron precursors allowing the solutions of forward and adjoint problems of static and dynamic states. The tool is applicable to both thermal and fast systems with hexagonal geometries. In the dynamic problems, the small stationary fluctuations of macroscopic cross sections are considered as noise sources, then the induced first-order noise is solved fully in the frequency domain. Numerical algorithms for solving the static and noise equations are implemented using finite differences for spatial discretization and a power iterative solution. A coarse-mesh finite difference technique for accelerating the convergence has been adopted. Verification calculations have been performed and compared to analytical solutions based on a two-dimensional homogeneous system with two energy groups and one group of delayed neutron precursors, in which pointlike perturbations of thermal absorption cross section at central and noncentral positions are considered as noise sources.
  •  
47.
  • Vidal-Ferrandiz, A., et al. (författare)
  • Neutronic Simulation of Fuel Assembly Vibrations in a Nuclear Reactor
  • 2020
  • Ingår i: Nuclear Science and Engineering. - : Informa UK Limited. - 0029-5639 .- 1943-748X. ; 194:11, s. 1067-1078
  • Tidskriftsartikel (refereegranskat)abstract
    • The mechanical vibrations of core internals such as fuel assemblies (FAs) cause oscillations in the neutron flux that require in some circumstances nuclear power plants to operate at a reduced power level. This work simulates and analyzes the changes of the neutron flux throughout a nuclear core due to the oscillation of a single FA without considering thermal-hydraulic feedback. The amplitude of the FA vibration is bounded to a few millimeters, and this implies the use of fine meshes and accurate numerical solvers due to the different scales of the problem. The results of the simulations show a main oscillation of the neutron flux with the same frequency as the FA vibration along with other harmonics at multiples of the vibration frequency much smaller in amplitude. Also, this work compares time domain analysis and frequency domain analysis of the mechanical vibrations. Numerical results show a close match between these two approaches for the fundamental frequency.
  •  
48.
  • Wallenius, Janne, 1968-, et al. (författare)
  • Application of burnable absorbers in an accelerator-driven system
  • 2001
  • Ingår i: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 137:1, s. 96-106
  • Tidskriftsartikel (refereegranskat)abstract
    • 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.
  •  
49.
  • Zylbersztejn, Florian, et al. (författare)
  • Calculation of the Neutron Noise Induced by Periodic Deformations of a Large Sodium-Cooled Fast Reactor Core
  • 2014
  • Ingår i: Nuclear Science and Engineering. - 0029-5639 .- 1943-748X. ; 177:2, s. 203-218
  • Tidskriftsartikel (refereegranskat)abstract
    • The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed.
  •  
50.
  • Jansson, Peter, 1971-, et al. (författare)
  • Gamma-Ray Spectroscopy Measurements of Decay Heat in Spent Nuclear Fuel
  • 2002
  • Ingår i: Nuclear Science and Engineering. - 0029-5639. ; 141:2, s. 129-139
  • Tidskriftsartikel (refereegranskat)abstract
    • A method for determining the residual thermal power in spent nuclear fuel using gamma-ray spectroscopy is suggested. It is based on the correlation between the residual power and the 137Cs activity, which is nearly linear for fuel with cooling times between 10 and 50 yr. Using available data of calorimetrically measured values of the decay heat in 69 boiling water reactor and pressurized water reactor spent-fuel assemblies resulted in agreement with a standard deviation of 3%.
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