Abstract
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) 
Fingerprint
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
Cite this
}
Development of a New Monte Carlo Reactor Physics code : Dissertation. / Leppänen, Jaakko.
Espoo : VTT Technical Research Centre of Finland, 2007. 241 p.Research output: Thesis › Dissertation › Monograph
TY  THES
T1  Development of a New Monte Carlo Reactor Physics code
T2  Dissertation
AU  Leppänen, Jaakko
PY  2007
Y1  2007
N2  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.
AB  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.
KW  reactor physics
KW  Monte Carlo method
KW  neutron transport codes
KW  PSG code
KW  deterministic reactor simulator codes
KW  homogenisation
KW  homogenised group constants
KW  nodal diffusion method
KW  LWR lattice calculations
M3  Dissertation
SN  9789513870188
T3  VTT Publications
PB  VTT Technical Research Centre of Finland
CY  Espoo
ER 