| 1. |
- Talamo, Alberto, et al.
(författare)
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Comparison of MCB and MONTEBURNS Monte Carlo burnup codes on a one-pass deep burn
- 2006
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:14-15, s. 1176-1188
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Tidskriftsartikel (refereegranskat)abstract
- Numerical applications implemented on the Monte Carlo method have developed in line with the increase of computer power; nowadays, in the field of nuclear reactor physics, it is possible to perform burnup simulations in a detailed 3D geometry and a continuous energy description by the Monte Carlo method; moreover, the required computing time can be abundantly reduced by taking advantage of a computer cluster. In this paper we focused on comparing the results of the two major Monte Carlo burnup codes, MONTEBURNS and MCB, when they share the same MCNP geometry, nuclear data library, core thermal power, and they apply the same refueling and shuffling schedule. While simulating a total operation time of the Gas Turbine-Modular Helium Reactor of 2100 effective full power days and a one-pass deep burn in-core fuel management schedule, we have found that the two Monte Carlo codes produce very similar results both on the criticality value of the core and the transmutation of the key actinides.
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| 2. |
- Dahlfors, Marcus, et al.
(författare)
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Neutron Cross Section Sensitivity for Minor Actinide Transmutation in Energy Amplifier Systems
- 2007
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 34:10, s. 824-835
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Tidskriftsartikel (refereegranskat)abstract
- The nuclear data sensitivity in 3D Monte Carlo burnup calculations of minor actinide transmutation in Energy Amplifier Systems is assessed. Ansaldo Nucleare's 80 MWth, Energy Amplifier Demonstration Facility (EADF) design serves as a technical and geometrical platform for the analysis. The accelerator-d riven EADF is a fast, subcritical system based on classical MOX-fuel technology and on molten lead-bismuth eutectic cooling. For Monte Carlo simulations, the state-of-the-art computer code package EA-MC, developed by C. Rubbia and his group at CERN, is utilised. The code offers treatment of both high-energy particle interactions and low-energy neutron transport with a sophisticated method based on a full Monte Carlo simulation, together with the option of employing different modern nuclear data libraries. In particular, the impact of nuclear data discrepancies on transmutation properties prediction with increasing exposure is examined. Monte Carlo simulation accelerator-driven systems transmutation burnup fast neutron spectrum minor actinides nuclear waste.
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| 3. |
- Degweker, Shashikant, 1956, et al.
(författare)
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Stochastic equations in the invariant imbedding formulation of particle transport
- 2009
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1108-1119
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Tidskriftsartikel (refereegranskat)abstract
- Invariant imbedding theory is an alternative formulation of particle transport theory. Although stochasticfoundations of invariant imbedding have been known from the beginnings, the method itself has so farexclusively been used for calculating first moments, i.e. expectations. The present paper attempts toset up a probability balance equation in the invariant imbedding approach from which equations forthe first and second order densities are derived. It is shown that only the equations for the first order densitiesare non-linear, while subsequent order densities obey linear equations. This is expected to considerablysimplify solution to those problems which involve second order density calculations whereinvariant imbedding techniques may be profitably used. Examples of such quantities are the varianceor correlations between particles detected at two different energies or angles or the higher momentsof the emitted multiplicity distribution such as the variance from a target bombarded by incident particles.One possible area of application of our equations is non-destructive estimation of fissile material bythe active neutron assay technique. Another area is the study of particle cascade development in sputteringand positron backscattering from surfaces. The approach is illustrated by a simple forward–backwardscattering model for these two problems.
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| 4. |
- Demaziere, Christophe, 1973, et al.
(författare)
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Identification and localization of absorbers of variable strength in nuclear reactors
- 2005
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:8, s. 812-842
-
Tidskriftsartikel (refereegranskat)abstract
- This paper investigates the possibility of localising a noise source of the type &DPRIME; absorber of variable strength&DPRIME; (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piece-wise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located.
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| 5. |
- Demaziere, Christophe, 1973, et al.
(författare)
-
On the possibility of the space-dependence of the stability indicator (decay ratio) of a BWR
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:12, s. 1305-1322
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Tidskriftsartikel (refereegranskat)abstract
- A model is proposed for the explanation of the space-dependence of the so-called decay ratio (DR) which is used to quantify the stability properties of boiling water reactors (BWRs). The study was prompted by the observation of a strongly space-dependent decay ratio in an instability event at the Swedish Forsmark-1 BWR. Prior to that event, the space-dependence of the DR was neither observed, nor assumed possible in the theoretical models of instability. The model proposed here is based on a previous suggestion by one of the authors on how to model the estimation of the DR in case of two different types of oscillations (instabilities) being present in the core simultaneously. The model was earlier only used in a space-independent form, but here its applicability is extended such that space-dependence of the oscillations is also accounted for, by using a noise simulator. The investigations show that the DR, as determined by the individual LPRMs (neutron detectors) at different positions, can be strongly space-dependent if at least two different oscillations with differing DR and space-dependence exist in the core simultaneously. The observed space-dependence of the DR in the Forsmark case can be reconstructed by the model.
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| 6. |
- Dufek, Jan, et al.
(författare)
-
An efficient parallel computing scheme for Monte Carlo criticality calculations
- 2009
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1276-1279
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Tidskriftsartikel (refereegranskat)abstract
- The existing parallel computing schemes for Monte Carlo criticality calculations suffer from a low efficiency when applied on many processors. We suggest a new fission matrix based scheme for efficient parallel computing. The results are derived from the fission matrix that is combined from all parallel simulations. The scheme allows for a practically ideal parallel scaling as no communication among the parallel simulations is required, and inactive cycles are not needed. (C) 2009 Elsevier Ltd. All rights reserved.
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| 7. |
- Dufek, Jan, et al.
(författare)
-
Fission matrix based Monte Carlo criticality calculations
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:8, s. 1270-1275
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Tidskriftsartikel (refereegranskat)abstract
- We have described a fission matrix based method that allows to cancel the inactive cycles in Monte Carlo criticality calculations. The fission matrix must be sampled in the course of the Monte Carlo calculation using a space mesh with sufficiently small zones as it causes the fission matrix be insensitive to errors in the initial fission source. The k(eff) and other quantities can be derived by means of the final fission matrix. The confidence interval for the k(eff) estimate can be conservatively determined via the variance in the fission matrix.
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| 8. |
- Dufek, Jan, et al.
(författare)
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Stability and convergence problems of the Monte Carlo fission matrix acceleration methods
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1648-1651
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Tidskriftsartikel (refereegranskat)abstract
- The Monte Carlo fission matrix acceleration methods aim at accelerating the convergence of the fission source in inactive cycles of Monte Carlo criticality calculations. In practice, however, these methods may corrupt the fission source, or slow down its convergence. These phenomena have not been completely understood so far. We demonstrate the convergence problems, and explain their reasons.
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| 9. |
- Dykin, Victor, 1985, et al.
(författare)
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Remark on the role of the driving force in BWR instability
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1544-1552
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Tidskriftsartikel (refereegranskat)abstract
- Simple models of BWR instability, used e.g. in understanding the role of the various oscillation modes inthe overall stability of the plant, assume that each oscillation mode can be described by a second ordersystem (a damped harmonic oscillator) driven by a white noise driving force. Change of the decay ratio(DR) of the observed signal is, as a rule, associated with the changing of the parameters of the dampedoscillator, mainly its damping coefficient, and is interpreted in terms of the change of the stability ofthe system. However, conceptually, one cannot exclude cases when the change of the response of a drivendamped oscillator is due to the change of the properties of the driving force. In this work we investigatethe effect of a non-white driving force on the behaviour of the system. A question of interest is howchanges of the spectrum of the driving force influence the observed autocorrelation function (ACF) ofthe resulting signal. Hence we calculate the response of a damped harmonic oscillator driven by anon-white driving force, corresponding to the reactivity effect of propagating density fluctuations intwo-phase flow. It is shown how in some special cases such a driving force, when interpreting the neutronnoise as if induced by a white noise driving source, can lead to an erroneous conclusion regarding thestability of the system. It is also concluded that in the practically interesting cases the effect of the coloureddriving force, arising from propagating density fluctuations, is negligible.
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| 10. |
- Gonzalez-Pintor, Sebastian, 1983, et al.
(författare)
-
High Order Finite Element Method for the Lambda modes problem on hexagonal geometry
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:9, s. 1450-1462
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Tidskriftsartikel (refereegranskat)abstract
- A High Order Finite Element Method to approximate the Lambda modes problem for reactors with hexagonal geometry has been developed. This method is based on the expansion of the neutron flux in terms of the modified Dubiner's polynomials on a triangular mesh. This mesh is fixed and the accuracy of the method is improved increasing the degree of the polynomial expansions without the necessity of remeshing. The performance of method has been tested obtaining the dominant Lambda modes of different 2D reactor benchmark problems. © 2009 Elsevier Ltd. All rights reserved.
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| 11. |
- Kitamura, Yasunori, et al.
(författare)
-
Calculation of higher moments of the neutron multiplication process in a time-varying medium
- 2007
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 34:5, s. 385-395
-
Tidskriftsartikel (refereegranskat)abstract
- The zero-power reactor noise theory in a steady neutron multiplying medium was extended recently to a medium randomly varying in time to bridge the fields of the zero-power and the power reactor noise. For a time-varying medium in which the transition probability randomly fluctuates, only the use of the probability generating function technique based on the forward master equation approach is practical. However, with the forward master equation approach, the treatment of the joint moments of the neutron number and the medium state leads to a closure problem. Recently, the capability of the moment calculation technique for such cases was extended such that the closure problem could be solved exactly. The present paper describes and demonstrates this closure-free moment calculation technique in a time-varying binary multiplying medium, in which the medium state has two possible realizations. In addition to the first two moments of the neutron number N alone (irrespective of the medium state η), the joint moments of Nn and ηm, i.e., , were also obtained in a compact form for n = 1, 2 and arbitrary values of m, without a closure assumption. It was found that, for even m values, the asymptotic values of Nn and ηm are uncorrelated, whereas, for odd m values, they are negatively correlated, namely, their covariance is less than zero. The first two moments of the neutron number theoretically obtained were verified by the Monte Carlo technique. A perfect agreement was found between the Monte Carlo and the theoretical solutions. The closure-free moment calculation technique demonstrated in the present paper is expected to be applicable to various other problems related to the birth-and-death process with fluctuations of the transition probability, in which a closure problem occurs.
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| 12. |
- Kitamura, Yasunori, et al.
(författare)
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Some properties of zero power neutron noise in a time-varying medium with delayed neutrons
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:9, s. 1621-1627
-
Tidskriftsartikel (refereegranskat)abstract
- The temporal evolution of the distribution of the number ofneutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to thecomplexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed.
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| 13. |
- Larsson, Viktor, 1984, et al.
(författare)
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Comparative study of 2-group P1 and diffusion theories for the calculation of the neutron noise in 1D 2-region systems
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:10, s. 1574-1587
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the neutron noise, i.e. the stationary fluctuations of the neutron flux around its meanvalue, is calculated in 2-group P1 and diffusion theories for a 2-region slab reactor using Green’sfunction technique. The applicability of diffusion theory for different types and locations of the perturbation,as well as different frequencies, is assessed. Material data, i.e. nuclear cross-sections and kineticparameters, representative of a Light Water Reactor (LWR) and of a Heavy Water Reactor (HWR),respectively, are used in this work. It is demonstrated that for practical situations in LWRs and HWRs,there is no significant advantage to use P1 theory since diffusion theory gives acceptable results. Thelargest deviations between the two formalisms are observed in regions of large gradients of the staticneutron flux, such as close to the reflector interface and close to the perturbation. Such observationsare in agreement with theoretical expectations. This study also indicates that neglecting the effect ofcross-section perturbation on the diffusion coefficient gives a rather small impact on the solution. Thisallows drastically simplifying the determination of the neutron noise. When using numericaltechniques for such a determination the memory requirements and computational effort can be significantlyreduced.
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| 14. |
- Mathews, T.S, et al.
(författare)
-
Integration of functional reliability analysis with hardware reliability : An application to safety grade decay heat removal system of Indian 500 MWe PFBR
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:4, s. 481-492
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Tidskriftsartikel (refereegranskat)abstract
- A passive system can fail either due to classical mechanical failure of components, referred to as hardware failure, or due to the failure of physical phenomena to fulfill the intended function, referred to as functional failure. In this paper a methodology is discussed for the integration of these two kinds of unreliability and applied to evaluate the integrated failure probability of the passive decay heat removal system of Indian 500 MWe prototype fast breeder reactor (PFBR). The probability of occurrence of various system hardware configurations is evaluated using the fault tree method and functional failure probabilities on the corresponding configurations are determined based on the overall approach reported in the reliability methods for passive system (RMPS) project. The variation of functional reliability with time, which is coupled to the probability of occurrence of various hardware system configurations is studied and incorporated in the integrated reliability analysis. It is observed that this consideration of the dependence of functional reliability on time will give significant advantages on system reliability. The integrated reliability analysis is also explained using an event tree. The impact of the provision for forced circulation in the primary circuit on functional reliability is also studied with this procedure and it is found that the forced circulation capability helps to bring down the total decay heat removal failure probability by lowering the peak temperatures after the reactor shut down.
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| 15. |
- Matsson, Ingvar, et al.
(författare)
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The shut-down of the Barseback 1 BWR : A unique opportunity to measure the power distribution in nuclear fuel rods
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:13, s. 1091-1101
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Tidskriftsartikel (refereegranskat)abstract
- Reactor poolside measurements of gamma radiation specific for the fission product La-140 (1596 keV) have been used for an experimental determination of axial power distributions in 55 nuclear fuel rods irradiated in the Barseback 1 BWR nuclear power plant. The measurements take advantage of the unique situation of a very short last reactor cycle of only three months due to the out-phasing of the reactor unit at November 30 1999. La-140 whose decay is controlled by the mother nuclide Ba-140 with the half-life 12.75 days reflects an average power distribution, representative for the latest weeks of core operation (in this case basically during November 1999). The measured intensities have been transformed into a 25 nodal representation to allow a precise and direct comparison with the corresponding calculated power distribution. The 55 rods were selected from two different fuel assemblies with average burn-ups of 1.9 and 9.7 MWd/ kgU, respectively (that is one fresh bundle and one slightly more than one cycle bundle). The stability and the linearity of the measurement system were evaluated. The linearity was checked using the two-source method. The stability was checked by recurrent measurements on a reference fuel rod. The results have been used in the validation of the pin power reconstruction model of Westinghouse 3D core simulator POLCA-7. The deviation between measured and calculated Ba-140 concentration (expressed as radial error) is typically a few percent on rod level. Results indicate that also Gd-rods are properly modelled over a broad range of conditions. It is indicated that predictions for fuel rods in their first month of operation are less accurate than for the rest of the rods.
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| 16. |
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| 17. |
- Pazsit, Imre, 1948, et al.
(författare)
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Calculation of the pulsed Feynman-alpha formulae and their experimental verification
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:9, s. 986-1007
-
Tidskriftsartikel (refereegranskat)abstract
- An effective method of calculating the pulsed Feynman-alpha formula for finite width pulses is introduced and applied in this paper. The method is suitable for calculating boththe deterministic and the stochastic Feynman-alpha formulae, while also being capable of treating various pulse shapes through very similar steps and partly identical formulae. In the paper both the deterministic and the stochastic cases are treated for square and Gaussianpulses. The solutions show a very good agreement with the results of currently performed experiments by some of the authors at the Kyoto University Critical Assembly (KUCA).The formulae obtained are also used for a quantitative evaluation of the prompt neutron decay constant from a large number of experiments made at the KUCA for a wide range of parameters such as subcritical reactivity, pulse repetition frequency and pulse width. The suitability of the formulae to determine the prompt neutron decay constant a by curve fitting to the measured data was investigated. It was found that, despite the larger deviation from thetraditional Feynman Y(T)-curves from the traditional ones with a constant source (i.e., larger ripples superimposed on a smooth curve), the stochastic pulsing method is superior to the deterministic one in that it yields the correct a value for all subcriticalities. The deterministicmethod also works fine for most cases, but its application is not so straightforward.
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| 18. |
- Persson, Carl-Magnus, et al.
(författare)
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Pulsed neutron source measurements in the subcritical ADS experiment YALINA-Booster
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:12, s. 2357-2364
-
Tidskriftsartikel (refereegranskat)abstract
- A subcritical zero-power source-driven coupled core, the YALINA-Booster. has been constructed for experimental investigations of neutron kinetics of source-driven systems. In this study, the reactivity of two subcritical configurations has been determined by the area ratio method. The prompt neutron decay constants have been evaluated through slope fitting of the prompt neutron decay as well as through the pulsed Rossi-alpha method. It is shown that the slope fitting method and the pulsed Rossi-alpha method give stable results whereas the area ratio method results show spatial dependence. The reasons for the spatial spread are addressed.
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| 19. |
- Seltborg, Per, et al.
(författare)
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Source efficiency as function of fuel and coolant in accelerator-driven systems
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:9, s. 829-832
-
Tidskriftsartikel (refereegranskat)abstract
- We have studied the efficiency of spallation neutron sources for different combinations of coolant and fuel in 80 MWth, sub-critical, cores. It has been found that the proton source efficiency, psi*, is reduced by 10% when switching coolant from helium to lead-bismuth eutectic. Substituting MOX fuel with an americium based fuel, results in another 10% reduction of psi*. The relatively high source efficiencies found for prototype accelerator-driven systems, using standard MOX fuel and helium coolant, may thus be difficult to achieve in future systems dedicated to the transmutation of higher actinides. Our results are in agreement with previous investigations of the dependence of the source efficiency on the selection of coolant.
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| 20. |
- Stålek, Mathias, 1978, et al.
(författare)
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Development and validation of a cross-section interface for PARCS
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:12, s. 2397-2409
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Tidskriftsartikel (refereegranskat)abstract
- This paper deals with the development of a cross-section interface for PARCS and its validation. The interface is used to feed PARCS with material constants for Light Water Reactors. These material constants are obtained from a CASMO-4 library file and the SIMULATE-3 code is then used to read this library file. This interface allows a dependency of the material constants on exposure and on instantaneous and history variables. Since the functionalization of the cross-sections in CASMO-4/SIMULATE-3 is different from the one in PARCS, the conversion of the material data from the CASMO-4/SIMULATE-3 formalism to the PARCS formalism is not trivial. As a first check of the proper conversion of the data by the interface, the cross-section files created by the interface were read by PARCS. The data were thereafter edited for all possible burnup, instantaneous and history parameters and compared to the original data used to create the files. After this successful verification, a benchmark between PARCS and plant measured data was carried out. For this benchmark a number of measurement sets from the Swedish Ringhals-3 pressurized water reactor were obtained. These data were measured during different cycles and at different core exposures. The spatial distribution of the instantaneous variables, the history variables and the exposure were calculated by SIMULATE-3 and used by PARCS to retrieve the actual three-dimensional distribution of the material data. The deviation of the effective multiplication factor keff from criticality was found to be within ±200 pcm. Both the measured axial and radial power profiles were adequately reproduced by the PARCS simulations, although some discrepancies with plant data need to be further investigated.
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| 21. |
- Talamo, Alberto, et al.
(författare)
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Adapting the Deep Burn In-Core Fuel Management Strategy for the Gas Turbine - Modular Helium Reactor to a Uranium-Thorium Fue
- 2005
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:16, s. 1750-1781
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Tidskriftsartikel (refereegranskat)abstract
- In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: U-235, which represents the 20% of the fresh uranium, U-233, which is produced by the transmutation of fertile Th-212, and Pu-239, which is produced by the transmutation of fertile U-238. In order to compensate the depletion of U-235 with the breeding of U-233 and Pu-239, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of U-235. At the same time, the amount of U-235 must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k(eff) and mass evolution, reaction rates, neutron flux and spectrum at the equilibrium of the fuel composition, highlights the features of a deep burn in-core fuel management strategy for a uranium-thorium fuel.
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| 22. |
- Talamo, Alberto
(författare)
-
Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium-thorium fuels
- 2007
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 34:1-2, s. 68-82
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Tidskriftsartikel (refereegranskat)abstract
- We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in U-235, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides Pu-240, U-238 and Th-232 and it requires the esteem of the Dancoff-Ginsburg factor for a pebble bed or prismatic core. The Dancoff-Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057-5692, 6.674-14485, 21.78-3472 eV for Pu-240, U-238 and Th-232, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TFUSO particles packing fraction from 100, 200 to 300 Pm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.
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| 23. |
- Talamo, Alberto
(författare)
-
Effects of the burnable poison heterogeneity on the long term control of excess of reactivity
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:9, s. 794-803
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Tidskriftsartikel (refereegranskat)abstract
- According to the different geometry shape, the theory of black burnable particles predicts that the evolution of the poison macroscopic absorption cross section is exponentially, quadratic or linear when the burnable poison is displaced in homogeneous distribution, microspheres or needlecylinders heterogeneous distributions, respectively. In the present studies, we took advantage of the Monte Carlo Continuous Energy Burnup Code MCB to verify the black burnable particles theory on the Gas Turbine - Modular Helium Reactor fuelled by military plutonium at the year the fuel reaches the equilibrium composition; we investigated 8 different burnable poisons, B, Cd, Er, Eu, Gd, Dy, Hf and Sm, in three different geometry configurations and we have found that the numerical results qualitatively match the theory predictions when burnable poisons are disposed in small particles.
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| 24. |
- Talamo, Alberto
(författare)
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Managing the Reactivity Excess of the Gas Turbine – Modular Helium Reactor by Burnable Poison and Control Rods
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:1, s. 84-98
-
Tidskriftsartikel (refereegranskat)abstract
- The gas turbine-modular helium reactor coupled to the deep burn in-core fuel management strategy offers the extraordinary capability to incinerate over 50% of the initial inventory of fissile material. This extraordinary feature, coming from an advanced and well tested fuel element design, which takes advantage of the TRISO particles technology, is maintained while the reactor is loaded with the most different types of fuels. In the present work, we assumed the reactor operating at the equilibrium of the fuel composition, obtained by a 6 years irradiation of light water reactor waste, and we investigated the effects of the introduction of the burnable poison and the control rods; we equipped the core with all the three types of control rods: operational, startup and shutdown ones. We employed as burnable poison natural erbium, due to the Er-167 increasing neutron capture microscopic cross-section in the energy range where the neutron spectrum exhibits the thermal peak; in addition, we utilized boron carbide, with 90% enrichment in 1013, as the absorption material of the control rods. Concerning the burnable poison studies, we focused on the k(eff) value, the Er-167 mass during burnup, the influence of modifying the radius of the BISO particles kernel and the fuel and moderator coefficients of temperature. Concerning the control rods studies, we investigated the reactivity worth, the changes in the neutron flux profile due to a partial insertion, the influence of modifying the radius of the BISO particles kernel and the beta(eff), at the beginning of the operation
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| 25. |
- Talamo, A., et al.
(författare)
-
MCB1C2 bug on thermal reactors
- 2006
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:7, s. 653-654
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Tidskriftsartikel (refereegranskat)
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| 26. |
- Talamo, Alberto, et al.
(författare)
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Performance of the Gas Turbine – Modular Helium Reactor fuelled with different types of fertile TRISO particles
- 2005
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 32:16, s. 1719-1749
-
Tidskriftsartikel (refereegranskat)abstract
- Preliminary studies have been performed on operation of the gas turbine-modular helium reactor (GT-MHR) with a thorium based fuel. The major options for a thorium fuel are a mixture with light water reactors spent fuel, mixture with military plutonium or with with fissile isotopes of uranium. Consequently, we assumed three models of the fuel containing a mixture of thorium with 239Pu, 233U or 235U in TRISO particles with a different kernel radius keeping constant the packing fraction at the level of 37.5%, which corresponds to the current compacting process limit. In order to allow thorium to act as a breeder of fissile uranium and ensure conditions for a self-sustaining fission chain, the fresh fuel must contain a certain quantity of fissile isotope at beginning of life; we refer to the initial fissile nuclide as triggering isotope. The small capture cross-section of 232Th in the thermal neutron energy range, compared to the fission one of the common fissile isotopes (239Pu, 233U and 235U), requires a quantity of thorium 25-30 times greater than that one of the triggering isotope in order to equilibrate the reaction rates. At the same time, the amount of the triggering isotope must be enough to set the criticality condition of the reactor. These two conditions must be simultaneously satisfied. The necessity of a large mass of fuel forces to utilize TRISO particles with a large radius of the kernel, 300 μm. Moreover, in order to improve the neutron economics, a fuel cycle based on thorium requires a low capture to fission ratio of the triggering isotope. Amid the common fissile isotopes, 233U, 235U and 239Pu, we have found that only the uranium nuclides have shown to have the suitable neutronic features to enable the GT-MHR to work on a fuel based on thorium.
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| 27. |
- Tampouratzi, Tatiani, 1965, et al.
(författare)
-
Non-invasive on-line two-phase flow regime identification employing artificial neural networks
- 2009
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 36:4, s. 464 - 469
-
Tidskriftsartikel (refereegranskat)abstract
- A novel non-invasive approach to the on-line identification of BWR two-phase flow regimes is investigated.The proposed approach receives neutron radiography images of coolant flow recordings as itsinput and performs feature extraction on each image via simple and directly computable statistical operators.The extracted features are subsequently used as inputs to an ensemble of self-organizing mapswhose outputs demonstrate swift and accurate classification of each image into its corresponding flowregime. The novelty of the approach lies in the use of the self-organizing map which generates the differentclasses by itself, according to feature similarity of the corresponding images; this contrasts traditionalartificial neural networks where the user has to define both the number of distinct classes as well as tosupply separate training vectors for each class.
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| 28. |
- Wallenius, Janne, 1968-, et al.
(författare)
-
Hafnium clad fuels for fast spectrum BWRs
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:1, s. 60-67
-
Tidskriftsartikel (refereegranskat)abstract
- We show that by use of hafnium cladding, a fast neutron spectrum is achievable in the top of uprated BWRs. Monte Carlo calculations have been made for Hf clad inert matrix nitride and low fertile MOX fuels, with fuel segments located in the upper part of an uprated BWR, where the coolant void fraction exceeds 70%. The nitride fuel results in the hardest neutron spectrum, but the low fertile MOX fuel still yields fission probabilities for even neutron number nuclides similar to those of sodium cooled reactors. The inert matrix nitride fuel configuration yields high burning rates, permitting to stabilise TRU inventories with less than 50% BWR cores of the here suggested type in the power park. The core with low fertile MOX fuel is less efficient, but still a zero net producer of TRU. Fuel and coolant temperature feedbacks are affected by introduction of absorbing elements in the fuel, but remain within acceptable ranges for the low fertile MOX fuel. Although control rod worths are reduced, shutdown margins are sufficient to ensure sub-criticality in cold conditions. From a materials point of view, the behaviour of hafnium clad MOX fuel would be similar to zircalloy clad MOX fuel already used extensively in nuclear industry. Thus, if dynamic stability of the core can be ensured, the here proposed fuel may be considered as a low cost solution for transmutation of minor actinides on industrial scale.
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| 29. |
- Westlén, Daniel, et al.
(författare)
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On TiN-particle fuel based helium-cooled transmutation systems
- 2006
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:16, s. 1322-1328
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Tidskriftsartikel (refereegranskat)abstract
- We have designed a sub-critical helium-cooled core with TiN-coated particle fuel, dedicated to the transmutation of minor actinides. The excellent neutronic properties of helium allows for a low plutonium fraction in the fuel, which yields a low reactivity swing, Delta k = 2600 pcm, for a burnup of 31.2%. Further the neutron spectrum is hard, limiting the buildup of Cm and Cf. The high burnup combined with a minor actinide burning rate of 355 kg/GWth year makes the present design an attractive transmutation system.
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| 30. |
- Willman, Christofer, et al.
(författare)
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A nondestructive method for discriminating MOX fuel from LEU fuel for safeguards purposes
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:9, s. 766-773
-
Tidskriftsartikel (refereegranskat)abstract
- Plutonium-rich mixed oxide fuel (MOX) is increasingly used in thermal reactors. However, spent MOX fuel could be a potential source of nuclear weapons material and a safeguards issue is therefore to determine whether a spent nuclear fuel assembly is of MOX type or of LEU (Low Enriched Uranium) type. In this paper, we present theoretical and experimental results of a study that aims to investigate the possibilities of using gamma-ray spectroscopy to determine whether a nuclear fuel assembly is of MOX or of LEU type. Simulations with the computer code ORIGEN-ARP have been performed where LEU and MOX fuel types with varying enrichment and burnup as well as different irradiation histories have been modelled. The simulations indicate that the fuel type determination may be achieved by using the intensity ratio Cs-134/Eu-154. An experimental study of MOX fuel of 14 x 14 PWR type and LEU fuel of both 15 x 15 and 17 x 17 type is also reported in this paper. The outcome of the experimental study support the conclusion that MOX fuel may be discriminated from LEU fuel by measuring the suggested isotopic ratio.
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| 31. |
- Willschütz, H. G., et al.
(författare)
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Recursively coupled thermal and mechanical FEM-analysis of lower plenum creep failure experiments
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:2, s. 126-148
-
Tidskriftsartikel (refereegranskat)abstract
- Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model (FEM) has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post-test calculations for the FOREVER test series representing the lower head RPV of a pressurised water reactor (PWR) in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. In this paper the differences between the results of a simple coupled and a recursive coupled FE-simulation are highlighted. Due to the thermal expansion at the beginning and the accumulating creep strain later on the shape of the melt pool and of the vessel wall are changing. Despite of the fact that these relative small geometrical changes take place relatively slowly over time, the effect on the temperature field is rather significant concerning the mechanical material behaviour and the resulting failure time. Assuming the same loading conditions, the change in the predicted failure time between the simple and the recursive coupled model is in the order of magnitude of the total failure time of the simple model. The comparison with results from the FOREVER-experiments shows that the recursive coupled model is closer to reality than the one-way coupled model.
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| 32. |
- Wright, Johanna, 1978, et al.
(författare)
-
Neutron kinetics in subcritical cores with application to the source modulation method
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 33:2, s. 149-158
-
Tidskriftsartikel (refereegranskat)abstract
- The subject of this paper is an investigation of the performance of the so-called source modulation technique for the measurement of reactivity in subcritical, source-driven cores. Methods of measuring reactivity by a single detector, including the source modulation method, are based on the assumption of point kinetic behaviour of the core. Deviations from point kinetic behaviour will lead to an inaccurate estimation of the reactivity. Hence, first, the conditions of point kinetic behaviour in subcritical source-driven cores are revisited. In addition to the known conditions for such behaviour, which have an analogy to those in critical cores, some additional cases are found which only exist in subcritical cores. Then the performance of the source modulation technique is investigated. It is found that the error of the method, originally thought to be due exclusively to the deviation of the local detector signal from the amplitude factor of point kinetics, remains finite and non-zero even in the limit of exact point kinetic behaviour (e.g., with low frequencies or deep subcriticalities). This is demonstrated and explained by analytical formulae. Some remedies for this shortcoming of the method are also suggested and discussed.
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| 33. |
- Zakova, Jitka, et al.
(författare)
-
Analysis of the reactivity coefficients of the advanced high-temperature reactor for plutonium and uranium fuels
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:5, s. 904-916
-
Tidskriftsartikel (refereegranskat)abstract
- The conceptual design of the advanced high-temperature reactor (AHTR) has recently been proposed by the Oak Ridge National Laboratory, with the intention to provide and alternative energy source for very high temperature applications. In the present study, we focused on the analyses of the reactivity coefficients of the AHTR core fueled with two types of fuel: enriched uranium and plutonium from the reprocessing of light water reactors irradiated fuel. More precisely, we investigated the influence of the outer graphite reflectors on the multiplication factor of the core, the fuel and moderator temperature reactivity coefficients and the void reactivity coefficient for five different molten salts: NaF, BeF2, LiF, ZrF4 and Li2BeF4 eutectic. In order to better illustrate the behavior of the previous parameters for different core configurations, we evaluated the moderating ratio of the molten salts and the absorption rate of the key fuel nuclides, which, of course, are driven by the neutron spectrum. The results show that the fuel and moderator temperature reactivity coefficients are always negative, whereas the void reactivity coefficient can be set negative provided that the fuel to moderator ratio is optimized (the core is undermoderated) and the moderating ratio of the coolant is large.
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| 34. |
- Zakova, Jitka, et al.
(författare)
-
Criticality assessment for prismatic high temperature reactors by fuel stochastic Monte Carlo modeling
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:5, s. 856-860
-
Tidskriftsartikel (refereegranskat)abstract
- Modeling of prismatic high temperature reactors requires a high precision description due to the triple heterogeneity of the core and also to the random distribution of fuel particles inside the fuel pins. On the latter issue, even with the most advanced Monte Carlo techniques, some approximation often arises while assessing the criticality level: first, a regular lattice of TRISO particles inside the fuel pins and, second, the cutting of TRISO particles by the fuel boundaries. We utilized two of the most accurate Monte Codes: MONK and MCNP, which are both used for licensing nuclear power plants in United Kingdom and in the USA, respectively, to evaluate the influence of the two previous approximations on estimating the criticality level of the Gas Turbine Modular Helium Reactor. The two codes exactly shared the same geometry and nuclear data library, ENDF/B, and only modeled different lattices of TRISO particles inside the fuel pins. More precisely, we investigated the difference between a regular lattice that cuts TRISO particles and a random lattice that axially repeats a region containing over 3000 non-cut particles. We have found that both Monte Carlo codes provide similar excesses of reactivity, provided that they share the same approximations.
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| 35. |
- Zinzani, Filippo, et al.
(författare)
-
Calculation of the eigenfunctions of the two-group neutron diffusion equation and application to modal decomposition of BWR instabilities
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 35:11, s. 2109-2125
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, numerical methods aiming at determining the eigenfunctions, their adjoint and the corresponding eigenvalues of the two-group neutron diffusion equations representing any heterogeneous system are investigated. First, the classical power iteration method is modified so that the calculation of modes higher than the fundamental mode is possible. Thereafter, the explicitly-restarted Arnoldi method, belonging to the class of Krylov subspace methods, is touched upon. Although the modified power iteration method is a computationally-expensive algorithm, its main advantage is its robustness, i.e. the method always converges to the desired eigenfunctions without any need from the user to set up any parameter in the algorithm. On the other hand, the Arnoldi method, which requires some parameters to be defined by the user, is a very efficient method for calculating eigenfunctions of large sparse system of equations with a minimum computational effort. These methods are thereafter used for off-line analysis of the stability of boiling water reactors in a two-dimensional representation of the core. Since several oscillation modes are usually excited (global and regional oscillations) when unstable conditions are encountered, the characterization of the stability of the reactor using for instance the Decay Ratio as a stability indicator might be difficult if the contribution from each of the modes are not separated from each other. Such a modal decomposition is applied to a stability test performed at the Swedish Ringhals-1 unit in September 2002, after the use of the Arnoldi method for pre-calculating the different eigenmodes of the neutron flux throughout the reactor. The modal decomposition clearly demonstrates the excitation of both the global and regional oscillations. Furthermore, such oscillations are found to be intermittent with a time-varying phase shift between the first and second azimuthal modes.
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| 36. |
- Arul, A.J, et al.
(författare)
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Reliability analysis of safety grade decay heat removal system of Indian prototype fast breeder reactor
- 2006
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549. ; 33:2, s. 180-188
-
Tidskriftsartikel (refereegranskat)abstract
- The 500MW Indian pool type Prototype Fast Breeder Reactor (PFBR), is provided with two independent and diverse Decay Heat Removal (DHR) systems viz., Operating Grade Decay Heat Removal System (OGDHRS) and Safety Grade Decay Heat Removal System (SGDHRS). OGDHRS utilizes the secondary sodium loops and Steam–Water System with special decay heat removal condensers for DHR function. The unreliability of this system is of the order of 0.1–0.01. The safety requirements of the present generation of fast reactors are very high, and specifically for DHR function the failure frequency should be less than 1E-7/ry. Therefore, a passive SGDHR system using four completely independent thermo-siphon loops in natural convection mode is provided to ensure adequate core cooling for all Design Basis Events. The very high reliability requirement for DHR function is achieved mainly with the help of SGDHRS. This paper presents the reliability analysis of SGDHR system. Analysis is performed by Fault Tree method using "CRAFT" software developed at Indira Gandhi Centre for Atomic Research. This software has special features for compact representation and CCF analysis of high redundancy safety systems encountered in nuclear reactors. Common Cause Failures (CCF) are evaluated by beta-factor method. The reliability target for SGDHRS arrived from DHR reliability requirement and the ultimate number of demands per year (7/y) on SGDHRS is that the failure frequency should be <=1.4E-8/de. Since it is found from the analysis that the unreliability of SGDHRS with identical loops is 5.2E-6/de and dominated by leak rates of components like AHX, DHX and sodium dump and isolation valves, options with diversity measures in important components were studied. The failure probability of SGDHRS for a design consisting of 2 types of diverse loops (Diverse AHX, DHX and sodium dump and isolation valves) is 2.1E-8/de, which practically meets the reliability requirement.
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| 37. |
- Ramakrishnan, M., et al.
(författare)
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Estimation of station blackout frequency for Indian fast breeder test reactor
- 2008
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549. ; 35:12, s. 2332-2337
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents the comparison of station blackout (SBO) frequency computed with approximate time averaged expressions for diesel generator unavailability and time dependent cutset evaluation method. It is found that the frequency of SBO is under predicted by a factor of 2 by approximate time averaged expressions for SBO durations of 8 h and 16 h. The time dependent cutset evaluation method is applied for offsite power feeder outage management by treating the change in SBO frequency when one of the feeders is taken out for maintenance for ‘n’ days, as the risk measure.
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| 38. |
- Willman, Christofer, et al.
(författare)
-
Nondestructive assay of spent nuclear fuel with gamma-ray spectroscopy
- 2006
-
Ingår i: Annals of Nuclear Energy. - 0306-4549. ; 33:5, s. 427-438
-
Tidskriftsartikel (refereegranskat)abstract
- An important issue in nuclear safeguards is to verify operator-declared data of spent nuclear fuel. Various techniques have therefore been assigned for this purpose. A nondestructive approach is to measure the gamma radiation from spent nuclear fuel assemblies. Using this technique, parameters such as burnup and cooling time can be calculated or verified. In this paper, we propose the utilization of gamma rays from 137Cs, 134Cs and 154Eu to determine the consistency of operator-declared information. Specifically, we have investigated to what extent irradiation histories can be verified. Computer simulations were used in order to determine limits for detecting small deviations from declared data. In addition, the technique has been experimentally demonstrated on 12 PWR fuel assemblies. A technique for determining burnup and cooling time for fuel assemblies where no operator-declared information is available is also presented. In such a case, the burnup could be determined with 1.6% relative standard deviation and the cooling time with 1.5%.
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| 39. |
- Grahn, Maria, 1963, et al.
(författare)
-
The role of biofuels for transportation in CO2 emission reduction scenarios with global versus regional carbon caps
- 2009
-
Ingår i: Biomass and Bioenergy. - : Elsevier BV. - 1873-2909 .- 0961-9534. ; 33:3, s. 360-371
-
Tidskriftsartikel (refereegranskat)abstract
- This study analyzes how international climate regimes affect cost-efficiency of fuel choices in the transportation sector. The analysis is carried out with a regionalized version of the Global Energy Transition model, GET-R 6.0. Two different carbon dioxide (CO2) reduction scenarios are applied, both meeting an atmospheric CO2 concentration target of 450 ppm by the year 2100. The first scenario, ‘‘global cap’’ (GC), uses a global cap on CO2 emissions, and global emissions trading is allowed. In the second scenario, ‘‘regional caps’’ (RC), industrialized regions start to reduce their CO2 emissions by 2010 while developing regions may wait several decades and emission reductions are not tradable across regions. In this second scenario, CO2 emissions are assumed to meet an equal per capita distribution of 1.0 tC/ capita, in all six regions, by 2040; emissions then follow a common reduction path, toward approximately 0.2 tC/capita by 2100. Three main results emerge from our analysis: (i) the use of biofuels in the industrialized regions is significantly higher in RC than in GC; (ii) the use of biofuels in RC actually increases the weaker (i.e., higher) the CO2 concentration target (up to 550 ppm); and (iii) biofuels never play a dominant role in the transportation sector. We find that biofuels may play a more important role in industrialized countries if these take on their responsibilities and reduce their emissions before developing countries start reducing their emissions, compared to the case in which all countries take action under a global cap and trade emission reduction regime.
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| 40. |
- Isaksson, Helena S., et al.
(författare)
-
Preanalytical aspects of quantitative TaqMan real-time RT-PCR : applications for TF and VEGF mRNA quantification
- 2006
-
Ingår i: Clinical Biochemistry. - Ottawa : Canadian soc. of clinical chemists. - 0009-9120 .- 1873-2933. ; 39:4, s. 373-377
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Tidskriftsartikel (refereegranskat)abstract
- OBJECTIVES: The present paper focuses on preanalytical aspects of tissue factor (TF) and vascular endothelial growth factor (VEGF) mRNA quantification: the choice of blood collection tubes and defining the time frame allowed before processing the sample. DESIGN AND METHODS: Blood was collected from healthy volunteers in K(3) EDTA tubes, CPT, endotoxin-free EndoTube tubes and in PAXgene tubes. Total RNA concentration was determined by absorbance readings at 260 nm with a GeneQuantII UV spectrophotometer. RNA quantity and quality were also determined by the Lab on a Chip technique (Agilent 2100 Bioanalyzer). Real-time RT-PCR assays were performed by the TaqMan technology. RESULTS: The more expensive PAXgene and CPT tubes and the Endo tubes did not give superior results from those obtained in inexpensive routine K(3) EDTA tubes. The PAXgene tubes preserved high molecular mass rRNA better than the other tubes. CONCLUSION: Both the PAXgene system and routine EDTA tubes are suitable for clinical purposes aimed at quantitation of mRNA for TF and VEGF. PAXgene yielded rRNA that was less degraded but had lower mRNA per microg extracted RNA. A time frame up to 24 h until sample processing is acceptable for TF and VEGF mRNA.
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| 41. |
- Hougner, Cajsa, et al.
(författare)
-
Economic valuation of a seed dispersal service in the Stockholm National Urban Park, Sweden
- 2006
-
Ingår i: Ecological Economics. - : Elsevier BV. - 0921-8009 .- 1873-6106. ; 59:3, s. 364-374
-
Tidskriftsartikel (refereegranskat)abstract
- Most economic valuation studies of species derive from stated preferences methods. These methods fail to take into account biodiversity values that the general public is not (made) informed about or has no experience with. Hence, production function (PF) and replacement cost (RC) approaches to valuation may be preferable in situations where species perform key life support functions in ecosystems, such as seed dispersal, pollination, or pest regulation. We conduct an RC analysis of the seed dispersal service performed by the Eurasian jay (Garrulus glandarius) in the Stockholm National Urban Park, Sweden. The park holds one of the largest populations of giant oaks in Europe, and the oak (Quercus robur and Quercus petrea) represents a keystone species in the hemiboreal forests. The primary objective was to estimate the number of seed-dispersed oak trees that resulted from jays and to determine the costs of replacing this service though human means. Results show that depending upon seeding or planting technique chosen, the RC per pair of jays in the park is SEK 35,000 (USD 4900) and SEK 160,000 (USD 22,500), respectively. Based on the park's aggregated oak forest-area, average RC for natural oak forest regeneration by jays is SEK 15,000 (USD 2100) to SEK 67,000 (USD 9400) per hectare, respectively. These estimates help motivating investments in management strategies that secure critical breeding and foraging habitats of jays, including coniferous forests and jay movement corridors. The analysis also illustrates the need for detailed ecological-economic knowledge in a PF or RC analysis. The continuous temporal and spatial oak dispersal service provided by jays holds several benefits compared to a man-made replacement of this service. PF and RC approaches are particularly motivated in cases of known functional ecological relationships, and critically important in estimating management measures where mobile link organisms and keystone species form key mutual relationships that generate high biodiversity benefits. In relation to obtained results, we discuss insights for conducting valuation studies on particular species.
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| 42. |
- Olsson, J., et al.
(författare)
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Applying climate model precipitation scenarios for urban hydrological assessment : a case study in Kalmar City, Sweden
- 2009
-
Ingår i: Atmospheric research. - : Elsevier BV. - 0169-8095 .- 1873-2895. ; 62:3, s. 364-375
-
Tidskriftsartikel (refereegranskat)abstract
- There is growing interest in the impact of climate change on urban hydrological processes. Such assessment may be based on the precipitation output from climate models. To date, the model resolution in both time and space has been too low for proper assessment, but at least in time the resolution of available model output is approaching urban scales. In this paper, 30-min precipitation from a model grid box covering Kalmar City, Sweden, is compared with high-resolution (tipping-bucket) observations from a gauge in Kalmar. The model is found to overestimate the frequency of low rainfall intensities, and therefore the total volume, but reasonably well reproduce the highest intensities. Adapting climate model data to urban drainage applications can be done in several ways but a popular way is the so-called Delta Change (DC) method. In this method, relative changes in rainfall characteristics estimated from climate model output are transferred to an observed rainfall time series, generally by multiplicative factors. In this paper, a version of the method is proposed in which these DC factors (DCFs) are related to the rainfall intensity level. This is achieved by calculating changes in the probability distribution of rainfall intensities and modelling the DCFs as a function of percentile. Applying this method in Kalmar indicated that in summer and autumn, high intensities will increase by 20-60% by year 2100, whereas low intensities remain stable or decrease. In winter and spring, generally all intensity levels increase similarly. The results were transferred to the observed time series by varying the volume of the tipping bucket to reflect the estimated intensity changes on a 30-min time scale. In an evaluation of the transformed data at a higher 5-min resolution, effects on the intensity distribution as well as single precipitation events were demonstrated. In particular, qualitatively different changes in peak intensity and total volume are attainable, which is required in light of expected future changes of the precipitation process and a step forward as compared with simpler DC approaches. Using the DC transformed data as input in urban drainage simulations for a catchment in Kalmar indicated an increase of the number of surface floods by 20-45% during this century.
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