Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor
-
P. N. V. Ha
Abstract
To cope with increasing spent fuel disposals and limited domestic spent fuel storages in Korea, the Korea Atomic Energy Research Institute has developed an advanced sodium-cooled fast reactor for TRU transmutation with an electricity output of 600 MWe (called the KALIMER-600 TRU burner). The design philosophy of the KALIMER-600 TRU burner concept is highly focused on inherent safety mechanisms, i.e., passive responses to abnormal and emergency conditions, and thereby minimizes the need for engineered safety systems. Accordingly, the main concern is on the sodium coolant void reactivity, a very important safety parameter of the KALIMER-600 TRU burner. This study was therefore performed to analyze the sodium void reactivity of the KALIMER-600 TRU burner to the finest resolution possible, e.g., contributions from any isotope in each core region. Such detailed analysis could be valuable and applicable to further optimization of the core passive safety characteristics against severe coolant voiding accident conditions.
Kurzfassung
Um dem Problem steigender Mengen abgebrannter Brennelemente und begrenzter eigener Lagermöglichkeiten in Korea gewachsen zu sein, hat das koreanische KAERI Forschungsinstitut einen fortgeschrittenen Natrium-gekühlten schnellen Reaktor für TRU Transmutation mit einer elektrischen Leistung von 600 MWe (den sogenannten KALIMER-600 TRU Brenner) entwickelt. Die Ausgestaltung des KALIMER-600 TRU Brennerkonzepts ist stark ausgerichtet auf inhärente Sicherheitsmechanismen und verringert dadurch die Notwendigkeit sicherheitstechnischer Anlagen. Dementsprechend liegt das Hauptinteresse bei der Void-Reaktivität des Natrium-Kühlmittels, einem wichtigen Sicherheitsparameter des KALIMER-600 TRU Brenners. Ziel dieser Studie war deshalb die Analyse der Natrium-Void-Reaktivität mit der bestmöglichen Auflösung. Eine solch detaillierte Analyse könnte auch für weitere Optimierungsmaßnahmen der passiven Sicherheitseigenschaften gegenüber schweren Kühlmittelverluststörfällen wichtig sein.
References
1 Generation IV International Forum: GIF R&D Outlook for Generation IV Nuclear Energy Systems, August 21, 2009Search in Google Scholar
2 Hahn, D. H. et al.: Advanced SFR design concepts and R&D activities. Nuclear Engineering and Technology41 (2009) 427-44610.5516/NET.2009.41.4.427Search in Google Scholar
3 Kim, Y. I. et al.: Preliminary Conceptual Design Report of Gen-IV SFR Demonstration Plant. KAERI/TR-4335/2011, Korea Atomic Energy Research Institute, 2011Search in Google Scholar
4 Kim, Y. I. et al.: Design concept of advanced sodium-cooled fast reactor and related R&D in Korea. Science and Technology of Nuclear Installations (2013), Article ID 290362, 1–1810.1155/2013/290362Search in Google Scholar
5 Dobbin, K. D. et al.: Evaluating the efficacy of a minor actinide burner. Global 93 Conference, Seattle, USA, September 199310.2172/10170117Search in Google Scholar
6 Hejzlar, P. et al.: Minor actinide burning in dedicated lead-bismuth cooled fast reactors. Proceedings of the 9th International Conference of Nuclear Engineering, Nice, France, April 8–12, 2001Search in Google Scholar
7 Hong, S. K.; Kim, S. J.; Kim, Y. I.: Annular fast reactor cores with low sodium void worth for TRU burning. Nuclear Technology162 (2008) 1-2510.13182/NT162-1-25Search in Google Scholar
8 Kim, T. K. et al.: Development of a perturbation code, PERT-K, for hexagonal core geometry. KAERI/TR-1194/98, Korea Atomic Energy Research Institute, 1998Search in Google Scholar
9 Jang, J. W.; Yoo, J.; Kim, Y. I.: Development of a perturbation theory module for triangular-z geometry. Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 10–11, 2007Search in Google Scholar
10 Ha, P. N.V.: Development of an isotope-wise reactivity module in the perturbation code PERT-K. Transactions of the Korean Nuclear Society Autumn Meeting, Kyeongju, Korea, October 25–26, 2012Search in Google Scholar
11 Derstine, K. L.: DIF3D: A Code to Solve One-, Two-, and Three-Dimensional Finite-Difference Diffusion Theory Problems. ANL-82-64, ANL, 198410.2172/7157044Search in Google Scholar
12 Macfarlane, R. E.: TRANSX-2: A Code for Interfacing MATXS Cross Section Libraries to Nuclear Transport Codes. LA-12312-MS, LANL, 1993Search in Google Scholar
13 Alcouffe, R. E.: User's Guide for TWODANT: A Code Package for Two-dimensional, Diffusion-accelerated, Neutron transport. LA-10049-M, LANL, 1990Search in Google Scholar
© 2014, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor
- Flow accelerated corrosion study in feeder pipes
- Analysis of the flow instability among channels of the OTSG in the naval craft NPP
- Experimental investigation of the MSFR molten salt reactor concept
- Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes
- Atomistic nano-scale 3D simulations about effects of Cr percentage on the molecular dynamics parameters of Fe-9–12% Cr alloys at fusion reactor temperature conditions
- On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation
- A finite volume approach to the problem of heat transfer in axisymmetric annulus geometry with internal heating element using local analytical solution techniques
- CO2 doped γ-irradiated hydroxyapatite for EPR dosimetry
- An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Decomposition analysis of the sodium void reactivity of the Korean sodium-cooled fast reactor
- Flow accelerated corrosion study in feeder pipes
- Analysis of the flow instability among channels of the OTSG in the naval craft NPP
- Experimental investigation of the MSFR molten salt reactor concept
- Investigation of neutronic behavior in a CANDU reactor with different (Am, Th, 235U)O2 fuel matrixes
- Atomistic nano-scale 3D simulations about effects of Cr percentage on the molecular dynamics parameters of Fe-9–12% Cr alloys at fusion reactor temperature conditions
- On an analytical evaluation of the flux and dominant eigenvalue problem for the steady state multi-group multi-layer neutron diffusion equation
- A finite volume approach to the problem of heat transfer in axisymmetric annulus geometry with internal heating element using local analytical solution techniques
- CO2 doped γ-irradiated hydroxyapatite for EPR dosimetry
- An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors
