Using MCNP for fusion neutronics

Dissertation

Frej Wasastjerna

Research output: ThesisDissertationCollection of Articles

1 Citation (Scopus)

Abstract

Any fusion reactor using tritium-deuterium fusion will be a prolific source of 14 MeV neutrons. In fact, 80% of the fusion energy will be carried away by these neutrons. Thus it is essential to calculate what will happen to them, so that such quantities as the tritium breeding ratio, the neutron wall loading, heat deposition, various kinds of material damage and biological dose rates can be determined. Monte Carlo programs, in particular the widely-used MCNP, are the preferred tools for this. The International Fusion Materials Irradiation Facility (IFMIF), intended to test materials in intense neutron fields with a spectrum similar to that prevailing in fusion reactors, also requires neutronics calculations, with similar methods. In some cases these calculations can be very difficult. In particular shielding calculations - such as those needed to determine the heating of the superconducting field coils of ITER or the dose rate, during operation or after shutdown, outside ITER or in the space above the test cell of IFMIF - are very challenging. The thick shielding reduces the neutron flux by many orders of magnitude, so that analog calculations are impracticable and heavy variance reduction is needed, mainly importances or weight windows. On the other hand, the shields contain penetrations through which neutrons may stream. If the importances are much higher or the weight windows much lower at the outer end of such a penetration than at the inner end, this may lead to an excessive proliferation of tracks, which may even make the calculation break down. This dissertation describes the author's work in fusion neutronics, with the main emphasis on attempts to develop improved methods of performing such calculations. Two main approaches are described: trying to determine near-optimal importances or weight windows, and testing the "tally source" method suggested by John Hendricks as a way of biasing the neutron flux in angle.
Original languageEnglish
QualificationDoctor Degree
Awarding Institution
  • Aalto University
Supervisors/Advisors
  • Santoro, Robert, Supervisor, External person
  • Iida, Hiromasa, Supervisor, External person
  • Fischer, Ulrich, Supervisor, External person
Award date19 Dec 2008
Place of PublicationEspoo
Publisher
Print ISBNs978-951-38-7129-1
Electronic ISBNs978-951-38-7130-7
Publication statusPublished - 2008
MoE publication typeG5 Doctoral dissertation (article)

Fingerprint

fusion
neutrons
fusion reactors
tritium
flux (rate)
shielding
penetration
materials tests
dosage
field coils
shutdowns
irradiation
deuterium
breakdown
analogs
damage
heat
heating
cells
energy

Keywords

  • fusion neutronics
  • MCNP
  • variance reduction
  • importances
  • weight windows
  • tally source method

Cite this

Wasastjerna, F. (2008). Using MCNP for fusion neutronics: Dissertation. Espoo: VTT Technical Research Centre of Finland.
Wasastjerna, Frej. / Using MCNP for fusion neutronics : Dissertation. Espoo : VTT Technical Research Centre of Finland, 2008. 132 p.
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isbn = "978-951-38-7129-1",
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publisher = "VTT Technical Research Centre of Finland",
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Wasastjerna, F 2008, 'Using MCNP for fusion neutronics: Dissertation', Doctor Degree, Aalto University, Espoo.

Using MCNP for fusion neutronics : Dissertation. / Wasastjerna, Frej.

Espoo : VTT Technical Research Centre of Finland, 2008. 132 p.

Research output: ThesisDissertationCollection of Articles

TY - THES

T1 - Using MCNP for fusion neutronics

T2 - Dissertation

AU - Wasastjerna, Frej

PY - 2008

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N2 - Any fusion reactor using tritium-deuterium fusion will be a prolific source of 14 MeV neutrons. In fact, 80% of the fusion energy will be carried away by these neutrons. Thus it is essential to calculate what will happen to them, so that such quantities as the tritium breeding ratio, the neutron wall loading, heat deposition, various kinds of material damage and biological dose rates can be determined. Monte Carlo programs, in particular the widely-used MCNP, are the preferred tools for this. The International Fusion Materials Irradiation Facility (IFMIF), intended to test materials in intense neutron fields with a spectrum similar to that prevailing in fusion reactors, also requires neutronics calculations, with similar methods. In some cases these calculations can be very difficult. In particular shielding calculations - such as those needed to determine the heating of the superconducting field coils of ITER or the dose rate, during operation or after shutdown, outside ITER or in the space above the test cell of IFMIF - are very challenging. The thick shielding reduces the neutron flux by many orders of magnitude, so that analog calculations are impracticable and heavy variance reduction is needed, mainly importances or weight windows. On the other hand, the shields contain penetrations through which neutrons may stream. If the importances are much higher or the weight windows much lower at the outer end of such a penetration than at the inner end, this may lead to an excessive proliferation of tracks, which may even make the calculation break down. This dissertation describes the author's work in fusion neutronics, with the main emphasis on attempts to develop improved methods of performing such calculations. Two main approaches are described: trying to determine near-optimal importances or weight windows, and testing the "tally source" method suggested by John Hendricks as a way of biasing the neutron flux in angle.

AB - Any fusion reactor using tritium-deuterium fusion will be a prolific source of 14 MeV neutrons. In fact, 80% of the fusion energy will be carried away by these neutrons. Thus it is essential to calculate what will happen to them, so that such quantities as the tritium breeding ratio, the neutron wall loading, heat deposition, various kinds of material damage and biological dose rates can be determined. Monte Carlo programs, in particular the widely-used MCNP, are the preferred tools for this. The International Fusion Materials Irradiation Facility (IFMIF), intended to test materials in intense neutron fields with a spectrum similar to that prevailing in fusion reactors, also requires neutronics calculations, with similar methods. In some cases these calculations can be very difficult. In particular shielding calculations - such as those needed to determine the heating of the superconducting field coils of ITER or the dose rate, during operation or after shutdown, outside ITER or in the space above the test cell of IFMIF - are very challenging. The thick shielding reduces the neutron flux by many orders of magnitude, so that analog calculations are impracticable and heavy variance reduction is needed, mainly importances or weight windows. On the other hand, the shields contain penetrations through which neutrons may stream. If the importances are much higher or the weight windows much lower at the outer end of such a penetration than at the inner end, this may lead to an excessive proliferation of tracks, which may even make the calculation break down. This dissertation describes the author's work in fusion neutronics, with the main emphasis on attempts to develop improved methods of performing such calculations. Two main approaches are described: trying to determine near-optimal importances or weight windows, and testing the "tally source" method suggested by John Hendricks as a way of biasing the neutron flux in angle.

KW - fusion neutronics

KW - MCNP

KW - variance reduction

KW - importances

KW - weight windows

KW - tally source method

M3 - Dissertation

SN - 978-951-38-7129-1

T3 - VTT Publications

PB - VTT Technical Research Centre of Finland

CY - Espoo

ER -

Wasastjerna F. Using MCNP for fusion neutronics: Dissertation. Espoo: VTT Technical Research Centre of Finland, 2008. 132 p.