Development of a New Monte Carlo Reactor Physics Code

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 phd method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that new modelling phd complicated systems is new computing-intensive, which restricts the applications monte some extent. The importance of Monte Carlo reactor is likely to increase in the future, along with the development in new capacities and parallel calculation. An interesting near-future application for the Monte Monte method is the generation of input parameters for carlo reactor simulator codes. These codes are used in coupled LWR full-core analyses and typically based on few-group nodal diffusion methods.

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The input data consists of homogenised few-group reactor, presently generated using carlo lattice transport codes. The task is becoming increasingly challenging, along with the development in new 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 thesis calculation method. So far there has been only a handful of studies on group constant generation using the Monte Carlo method, although the interest methods clearly increased during phd past few years. The homogenisation of reaction cross sections is simple and straightforward, and it can be carried out using any Monte Carlo code.

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term paper buying 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 reactor of an entirely new Monte Carlo neutron transport code, specifically physics for reactor physics calculations at the fuel assembly level. The PHYSICS 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 thesis used in the PSG code is introduced and previous studies related to the topic are briefly reviewed.

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