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Author: S. Podgor
Publisher:
ISBN:
Category : Naval research
Languages : en
Pages : 24
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Book Description
Author: S. Podgor
Publisher:
ISBN:
Category : Naval research
Languages : en
Pages : 24
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Book Description
Author: Jaakko Leppänen
Publisher:
ISBN: 9789513870188
Category : Monte Carlo method
Languages : en
Pages : 236
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Book Description
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation. An interesting near-future application for the Monte Carlo method is the generation of input parameters for deterministic reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on fewgroup nodal diffusion methods. The input data consists of homogenised fewgroup constants, presently generated using deterministic lattice transport codes. The task is becoming increasingly challenging, along with the development in nuclear technology. Calculations involving high-burnup fuels, advanced MOX technology and next-generation reactor systems are likely to cause problems in the future, if code development cannot keep up with the applications. A potential solution is the use of Monte Carlo based lattice transport codes, which brings all the advantages of the calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest has clearly increased during the past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code. Some of the parameters, however, require the use of special techniques that are usually not available in general-purpose codes. The main problem is the calculation of neutron diffusion coefficients, which have no continuous-energy counterparts in the Monte Carlo calculation. This study is focused on the development of an entirely new Monte Carlo neutron transport code, specifically intended for reactor physics calculations at the fuel assembly level. The PSG code is developed at VTT Technical Research Centre of Finland and one of the main applications is the generation of homogenised group constants for deterministic reactor simulator codes. The theoretical background on general transport theory, nodal diffusion calculation and the Monte Carlo method are discussed. The basic methodology used in the PSG code is introduced and previous studies related to the topic are briefly reviewed. PSG is validated by comparison to reference results produced by MCNP4C and CASMO-4E in infinite two-dimensional LWR lattice calculations. Group constants generated by PSG are used in ARES reactor simulator calculations and the results compared to reference calculations using CASMO-4E data.
Author: Aubey Rotenberg
Publisher:
ISBN:
Category : Monte Carlo method
Languages : en
Pages : 42
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Book Description
Author: Magdi M. H. Ragheb
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 580
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Book Description
Author: Richard L. Morin
Publisher: CRC Press
ISBN: 1000005933
Category : Medical
Languages : en
Pages : 244
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Book Description
First Published in 1988, this book offers a full exploration into the applications of the Monte Carlo Simulation. Carefully compiled and filled with a vast repertoire of notes, diagrams, and references this book serves as a useful reference for Students of Radiology, and other practitioners in their respective fields.
Author: I. Lux
Publisher: CRC Press
ISBN: 1351091735
Category : Science
Languages : en
Pages : 492
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Book Description
With this book we try to reach several more-or-less unattainable goals namely: To compromise in a single book all the most important achievements of Monte Carlo calculations for solving neutron and photon transport problems. To present a book which discusses the same topics in the three levels known from the literature and gives us useful information for both beginners and experienced readers. It lists both well-established old techniques and also newest findings.
Author: Teodor Roşescu
Publisher:
ISBN:
Category :
Languages : en
Pages : 49
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Book Description
Author: J. R. Streetman
Publisher:
ISBN:
Category : Monte Carlo method
Languages : en
Pages : 86
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Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
The coarse mesh finite difference (CMFD) formulation has been applied to Monte Carlo (MC) simulations in order to mitigate the issue of large real variances of pin power tallies in full-core problems. In this work, a parallelized multigroup (MG) two-dimensional (2-D) MC code named PRIDE (Probabilistic Reactor Investigation with Discretized Energy), which is capable of handling lattices of square pin cells within which circular substructures can be modeled, has been developed as a tool for the investigations of the new method. In this code, a scheme to construct a CMFD linear system is based on the MC tallies of coarse mesh average fluxes and the net currents at coarse mesh interfaces. These tallies are accumulated over the MC cycles to get more stable CMFD solutions which are used for feedback to MC fission source distribution (FSD). The feedback scheme in this code employs a weight adjustment of fission source neutrons for the next MC cycle that is to reflect the global CMFD FSD into the MC FSD. The performance of CMFD feedback has been investigated in terms of the number of inactive cycles required for the convergence of FSD and also the reduction of real variances of local property tallies in active cycles. The applications to 2-D multigroup full-core pressurized water reactor problems have demonstrated that the MC FSD converges considerably faster and the real variances of pin powers are smaller by a factor of 4 with CMFD FSD feedback. It is also noted that the large real variances of pin powers are caused mainly by the global assembly-wise fluctuations of power distributions in a large core rather than local fluctuations.
Author: H. Zaidi
Publisher: CRC Press
ISBN: 1000687686
Category : Medical
Languages : en
Pages : 441
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Book Description
Therapeutic Applications of Monte Carlo Calculations in Nuclear Medicine examines the applications of Monte Carlo (MC) calculations in therapeutic nuclear medicine, from basic principles to computer implementations of software packages and their applications in radiation dosimetry and treatment planning. With chapters written by recognized authorit
