Development of a New Monte Carlo Reactor Physics code: Dissertation

Research output: ThesisDissertationMonograph

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 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 few-group nodal diffusion methods. The input data consists of homogenised few-group 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 byMCNP4C 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.
Original languageEnglish
QualificationDoctor Degree
Awarding Institution
  • Aalto University
Supervisors/Advisors
  • Vanttola, Timo, Supervisor
  • Kotiluoto, Petri, Supervisor
Award date18 Jun 2007
Place of PublicationEspoo
Publisher
Print ISBNs978-951-38-7018-8
Electronic ISBNs978-951-38-7019-5
Publication statusPublished - 2007
MoE publication typeG4 Doctoral dissertation (monograph)

Fingerprint

reactor physics
reactors
simulators
Monte Carlo method
neutrons
Advanced Reconn Electric Spacecraft
radiation shielding
Finland
transport theory
homogenizing

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

Leppänen, J. (2007). Development of a New Monte Carlo Reactor Physics code: Dissertation. Espoo: VTT Technical Research Centre of Finland.
Leppänen, Jaakko. / Development of a New Monte Carlo Reactor Physics code : Dissertation. Espoo : VTT Technical Research Centre of Finland, 2007. 241 p.
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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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Development of a New Monte Carlo Reactor Physics code : Dissertation. / Leppänen, Jaakko.

Espoo : VTT Technical Research Centre of Finland, 2007. 241 p.

Research output: ThesisDissertationMonograph

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 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 few-group nodal diffusion methods. The input data consists of homogenised few-group 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 byMCNP4C 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.

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 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 few-group nodal diffusion methods. The input data consists of homogenised few-group 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 byMCNP4C 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.

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 - 978-951-38-7018-8

T3 - VTT Publications

PB - VTT Technical Research Centre of Finland

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Leppänen J. Development of a New Monte Carlo Reactor Physics code: Dissertation. Espoo: VTT Technical Research Centre of Finland, 2007. 241 p.