Showing papers in "Annals of Nuclear Energy in 2006"

TL;DR: In this paper, the linear chain method of solving Bateman equations for nuclear transmutation in derivation of the general solution for linear chain with repeated transitions and thus elimination of existing numerical problems is discussed.

TL;DR: A sensitivity and uncertainty study has been performed to evaluate the impact of neutron cross-section uncertainty on the most significant integral parameters related to the core and fuel cycle, and gives guidelines in order to define the most appropriate and effective strategy for data uncertainty reduction.

TL;DR: A methodology to identify basic events in the risk assessment that have the potential to change the decision and are known to have significant model uncertainties that have been identified in the literature as generally important is presented.

TL;DR: In this paper, a nondestructive approach is proposed to verify operator-declared data of spent nuclear fuel data of a nuclear power plant in order to verify its safety.

TL;DR: In this paper, the Karhunen-Loeve procedure and the associated polynomial chaos expansion have been employed to solve a simple first order stochastic differential equation which is typical of transport problems.

TL;DR: In this article, the authors consider the Cognitive Reliability and Error Analysis Method (CREAM) and propose some developments with respect to a systematic procedure for computing probabilities of action failure.

TL;DR: In this article, a fuel rod design for high power density supercritical water-cooled fast reactor was conducted with mixed-oxide (MOX) fuel and stainless steel (SUS304) cladding under the limiting cladding surface temperature of 650°C.

TL;DR: New algorithms based on the EDAs approach are applied to maximize the effective multiplication factor of the CONSORT research reactor of Imperial College London to improve the optimisation of nuclear reactor fuel management optimisation problems.

TL;DR: In this paper, an asymptotically exact method for the calculation of the effective (i.e., adjoint weighted) neutron generation time Λeff was proposed for continuous energy simulations.

TL;DR: The principles of database and Web application development used at NNDC are described and examples of nuclear structure, nuclear reaction and bibliographical database applications along with a number of nuclear science tools and codes are presented.

TL;DR: The fault diagnostic problem is tackled within a neuro-fuzzy approach to pattern classification, aimed at obtaining not only a high rate of correct classification but also a transparent classification model, i.e., readable and easily interpretable from the physical point of view.

TL;DR: In this article, the authors investigate the RELAP5/3.2 system code capabilities in predicting phenomena that could be encountered under abnormal research reactor's operating conditions, such as flow excursion tests performed at the Oak Ridge National Laboratory thermal hydraulic test loop.

TL;DR: In this article, the authors compare the results of a simple coupled and a recursive coupled finite element model (FE-simulation) for the FOREVER test series of a PWR in the geometrical scale of 1:10.

TL;DR: In this article, the results of steam explosion experiments using molten material consisting of UO 2 and ZrO 2 mixture, which is called corium, to simulate a prototypic steam explosion in a nuclear reactor during a postulated severe accident were discussed.

TL;DR: The 500MW Indian pool type Prototype Fast Breeder Reactor (PFBR) as discussed by the authors was provided with two independent and diverse Decay Heat Removal (DHR) systems viz., operating grade decay heat removal system (OGDH) and decomposition heat removal (DSHR) systems.

TL;DR: In this paper, an investigation of the thermal hydraulic characteristics and the natural circulation performance in the passive residual heat removal system (PRHRS) for an integral type reactor have been carried out using the VISTA facility and the calculated results using the MARS code, which is a best estimate system analysis code have been compared with the experimental results.

TL;DR: The results showed that the RBF networks produced slightly better results than the MLP networks, but, for practical purposes, both can be considered of similar accuracy.

TL;DR: In this article, a Monte Carlo scheme is presented to simulate particles going across an interface separating two layers of a medium characterized by different physical properties, together with an analytical formulation of the same problem, for both normal diffusive and subdiffusive regimes.

TL;DR: In this paper, measurements in the CROCUS reactor at EPFL, Lausanne, are reported for the critical water level and the inverse reactor period for several different sets of delayed supercritical conditions.

TL;DR: In this article, the integration scheme along a trajectory is discussed from the point of view of the method of characteristics, and the analogy with the discrete ordinates method in slab geometry is highlighted along with the practical limitation in transposing high-order S N schemes to a trajectory-based method.

TL;DR: The experimental data concerning the prompt fission neutron number as function of the fission fragment mass (currently named "sawtooth" data) when they exist, allow a more refined verification of the point by point model and in the same time the validation of the methods used to determine the model parameters corresponding to the fissure fragment pairs covering the entire fission range as discussed by the authors.

TL;DR: In this paper, a space asymptotic model for the study of pulsed experiments in neutron multiplying systems is presented, where the Laplace transformed one-group transport equation is derived.

TL;DR: The proposed approach can evaluate the error detection coverage and the fault-tolerance of DPPS in NPPs and uses C++ based hardware description instead of a board with integrated circuit components.

TL;DR: In this paper, the authors presented new ENDF-6 formatted nuclear data libraries for 54 Fe, 56 Fe, 57 Fe, and 58 Fe, for incident neutrons and protons between 0 and 200

TL;DR: In this article, a genetic algorithm was applied to design the optimal enriched gadolinium poisons in a reference PWR TMI-1 core, and the break-even point for unit Gd-155 enrichment cost was determined to be around ∼$30/gram-Gd155.

TL;DR: In this article, the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated, and four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal recycling.

TL;DR: In this paper, a 10MW IAEA MTR research reactor was subjected to unprotected reactivity insertion (step and ramp) for both low and high-enriched fuel with continuous reactivity feedback due to coolant and fuel temperature effects.

TL;DR: In this article, 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 are compared.

TL;DR: In this article, an equation that can predict the variance of nuclide number densities after Monte Carlo burn-up calculations was proposed, and the results revealed that the effects of error propagation on the Monte Carlo calculations of 8 × 8 BWR fuel assembly are low up to 60 GWd/t.

TL;DR: An efficient and a practical genetic algorithm (GA) tool was developed and applied successfully to Burnable Poison (BP) placement optimization problem in the reference Three Mile Island-1 (TMI-1) core.