A study on the neutronic performance of the ARIES-RS fusion reactor with various coolants bearing nuclear fuel
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M. Übeyli
Abstract
Heavy metal molten salts have a good potential to be used in fusion reactors due to their capability of increasing fusion power and producing fissile fuel. In this study, the effect of various molten salts bearing nuclear fuel (99.5 % Flibe + 0.5 % Weapon grade (WG) PuF4, 99 % Flibe + 1 % WG PuF4, 99.5 % Flibe + 0.5 % Reactor Grade (RG) PuF4, 99 % Flibe + 1 % RG PuF4, 75 % LiF + 25 %ThF4, 75 % LiF + 24 % ThF4 + 1 % 233UF4 and 75 % LiF + 23 %ThF4 + 2 % 233UF4) on the neutronic performance of a magnetic fusion reactor called ARIES-RS was investigated. Neutron transport calculations were performed with the help of the code, Scale4.3. Numerical results showed that a significant enhancement on the neutronic performance was achieved with respect to the energy multiplication and fissile fuel breeding by keeping the sufficient tritium for the (DT) driver compared to the pure ARIES-RS fusion reactor for all investigated coolants except for the salts of 75 % LiF + 25 % ThF4 and 75 % LiF + 24 % ThF4 + 1 % 233UF4. Since these two salts did not satisfy tritium self-sufficiency.
Kurzfassung
Geschmolzene Schwermetallsalze haben wegen ihrer Fähigkeit die Fusionsenergie zu erhöhen und spaltbaren Brennstoff zu erzeugen gute Vorraussetzungen bei Fusionsreaktoren zum Einsatz zu kommen. In der vorliegenden Arbeit wird der Einfluss verschiedener Brennstoff-haltiger geschmolzener Salze (99,5 % Flibe + 0,5 % Weapon grade (WG) PuF4, 99 % Flibe + 1 % WG PuF4, 99,.5 % Flibe + 0,5 % Reactor Grade (RG) PuF4, 99 % Flibe + 1 % RG PuF4, 75 % LiF + 25 % ThF4, 75 % LiF + 24 % ThF4 + 1 % 233UF4 and 75 % LiF + 23 % ThF4 + 2 % 233UF4) auf das neutronenphysikalische Verhalten eines magnetischen Fusionsreaktors, ARIES-RS, untersucht. Neutronentransportrechnungen wurden durchgeführt mit Hilfe des Codes Scale4.3. Die numerischen Ergebnisse zeigen, dass eine signifikante Erhöhung des neutronenphysikalische Verhaltens in Bezug auf die Energiemultiplikation und die Erzeugung spaltbaren Brennstoffs erreicht wird durch Aufbewahrung von ausreichend Tritium für die (DT) Treiber verglichen mit dem reinen ARIES-RS Fusionsreaktor für alle untersuchten Kühlmittel, außer für die Salze von 75 % LiF + 25 % ThF4 und 75 % LiF + 24 % ThF4 + 1 % 233UF4. Diese beiden Salze erfüllen nicht die Selbstversorgung mit Tritium.
References
1 Tillack, M. S.: Engineering design of the ARIES-RS power plant. Fusion Engineering and Design41 (1998) 49110.1016/S0920-3796(97)00161-0Search in Google Scholar
2 Najmabadi, F.: The ARIES Team, Overview of ARIES-RS tokamak fusion power plant. Fusion Engineering and Design41 (1998) 365Search in Google Scholar
3 Conn, R. W.; Najmabadi, F.; Tillack, M.S.: Prospects and issues for commercial fusion power systems. Fusion Engineering and Design41 (1998) 33710.1016/S0920-3796(97)00152-XSearch in Google Scholar
4 El-Guebaly, L. A.: The ARIES Team, Overview of ARIES-RS neutronics and radiation shielding: Key issues and main conclusions. Fusion Engineering and Design38 (1997) 139Search in Google Scholar
5 Steiner, D.; El-Guebaly, L.; Herring, S.et al.: ARIES-RS safety design and analysis. Fusion Engineering and Design38 (1997) 18910.1016/S0920-3796(97)00116-6Search in Google Scholar
6 Şahin, S.; Yalçın, Ş.; Şahin, H. M.; Übeyli, M.: Neutronic investigation of a hybrid version of the ARIES-RS fusion reactor. Annals of Nuclear Energy30 (2003) 245Search in Google Scholar
7 Übeyli, M.: Utilization of ceramic uranium fuels in ARIES-RS fusion reactor. Journal of Fusion Energy23 (2004) 41Search in Google Scholar
8 Übeyli, M.: Neutronic analysis of ARIES-RS fusion reactor fueled with thorium. Energy Conversion and Management47 (2006) 322Search in Google Scholar
9 IAEA, Potential of thorium based fuel cycles to constrain plutonium and reduce long lived waste toxicity, International Atomic Energy Agency, Vienna, Austria, IAEA-TECDOC-1349 (2003)Search in Google Scholar
10 Jordan, W. C.; Bowman, S. M.: Scale Cross-Section Libraries, NUREG/CR-0200, Revision 5, 3, section M4, ORNL/NUREG/CSD-2/V3/R5, Oak Ridge National Laboratory (1997)Search in Google Scholar
11 Greene, N. M.; Petrie, L. M.: XSDRNPM, A One-Dimensional Discrete-Ordinates Code for Transport Analysis, NUREG/CR-0200, Revision 5, 2, Section F3, ORNL/NUREG/CSD-2/V2/R5, Oak Ridge National Laboratory (1997)Search in Google Scholar
12 Şahin, S.: Radiation Shielding Calculations for Fast Reactors (in Turkish), Gazi University, Publication # 169, Faculty of Science and Literature, Publication number 22, Ankara, Turkey (1991)Search in Google Scholar
13 Greene, N. M.: BONAMI, Resonance Self-Shielding by the Bondarenko Method, NUREG/CR-0200, Revision 5, 2, section F1, ORNL/NUREG/CSD-2/V2/R5, Oak Ridge National Laboratory (1997)Search in Google Scholar
14 Greene, N. M.; Petrie, L. M.; Westfall, R. M.: NITAWL-II, Scale System Module for Performing Resonance Shielding and Working Library Production, NUREG/CR-0200, Revision 5, 2, Section F2, ORNL/NUREG/CSD-2/V2/R5, Oak Ridge National Laboratory (1997)Search in Google Scholar
15 Landers, N. F.; Petrie, L. M.: CSAS, Control Module for Enhanced Criticality Safety Analysis Sequences, NUREG/CR-0200, Revision 5, 1, Section C4, ORNL/NUREG/CSD-2/V1/R5, Oak Ridge National Laboratory (1997)Search in Google Scholar
16 Yapıcı, H.: XSCALC for Interfacing Output of XSDRN to Calculate Integral Neutronic Data, Erciyes University, Kayseri, Turkey (2001)Search in Google Scholar
© 2008, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Application of the HN method to radiative heat transfer for a non-conservative slab
- Application of the spectral Green's function method for numerically solving discrete ordinates problems in cylindrical geometry
- Chebyshev polynomial (TN) approximation to neutron transport theory and application to the critical slab problem
- Spherical albedo problem for isotropic scattering
- The slab albedo and criticality problem for the quadratic scattering kernel with the HN method
- An analytical solution for the time-dependent SN transport equation in a slab
- Deterministic calculation of grey Dancoff factors in cluster cells with cylindrical outer boundaries
- New calculations of excitation functions of some positron emitting and single photon emitting radioisotopes
- Prediction of the time-dependent failure rate for normally operating components taking into account the operational history
- Assessment of FEMAXI and TESPA-ROD codes for modelling of BDBA in RBMK-1500
- A study on the neutronic performance of the ARIES-RS fusion reactor with various coolants bearing nuclear fuel
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Application of the HN method to radiative heat transfer for a non-conservative slab
- Application of the spectral Green's function method for numerically solving discrete ordinates problems in cylindrical geometry
- Chebyshev polynomial (TN) approximation to neutron transport theory and application to the critical slab problem
- Spherical albedo problem for isotropic scattering
- The slab albedo and criticality problem for the quadratic scattering kernel with the HN method
- An analytical solution for the time-dependent SN transport equation in a slab
- Deterministic calculation of grey Dancoff factors in cluster cells with cylindrical outer boundaries
- New calculations of excitation functions of some positron emitting and single photon emitting radioisotopes
- Prediction of the time-dependent failure rate for normally operating components taking into account the operational history
- Assessment of FEMAXI and TESPA-ROD codes for modelling of BDBA in RBMK-1500
- A study on the neutronic performance of the ARIES-RS fusion reactor with various coolants bearing nuclear fuel
