Abstract
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 computingintensive, 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 nearfuture application for the Monte
Carlo method is the generation of input parameters for
deterministic reactor simulator codes. These codes are
used in coupled LWR fullcore 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 highburnup fuels, advanced MOX
technology and nextgeneration 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
generalpurpose codes. The main problem is the
calculation of neutron diffusion coefficients, which have
no continuousenergy 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 byMCNP4C and CASMO4E in
infinite twodimensional LWR lattice calculations. Group
constants generated by PSG are used in ARES reactor
simulator calculations and the results compared to
reference calculations using CASMO4E data.
Original language  English 

Qualification  Doctor Degree 
Awarding Institution 

Supervisors/Advisors 

Award date  18 Jun 2007 
Place of Publication  Espoo 
Publisher  
Print ISBNs  9789513870188 
Electronic ISBNs  9789513870195 
Publication status  Published  2007 
MoE publication type  G4 Doctoral dissertation (monograph) 
Keywords
 reactor physics
 Monte Carlo method
 neutron transport codes
 PSG code
 deterministic reactor simulator codes
 homogenisation
 homogenised group constants
 nodal diffusion method
 LWR lattice calculations
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Leppänen, J. (2007). Development of a New Monte Carlo Reactor Physics code: Dissertation. VTT Technical Research Centre of Finland. http://www.vtt.fi/inf/pdf/publications/2007/P640.pdf