Stability analysis of the Korean prototype Generation-IV sodium-cooled fast reactor using linear frequency domain approach
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S. J. Kim
, P. N. V. Ha , J. Y. Lim , D. H. Hahn und C. M. Kang
Abstract
The Korea Atomic Energy Research Institute (KAERI) has been developing the 150 MWe Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR). The design concept is highly based on passive safety mechanisms, minimizing the need for engineered safety systems. Presently, it is of primary importance to assure the reactor dynamics and stability against small reactivity disturbances under power operating conditions. KAERI has therefore developed the NuSTAB code for stability analysis of the PGSFR. In NuSTAB, the neutron-kinetic and thermal-hydraulic coupling equations are linearized to form the characteristic equation, which is solved as a generalized eigenvalue problem for determining the decay ratio, an indicator of the system stability. In this paper, the stability of the PGSFR was analyzed by applying the point kinetic and spatial kinetic options in the NuSTAB code. System responses to temperature feedbacks including the Doppler effect, thermal expansion, coolant density change, and overall feedback were studied. The results indicate that the initial U and final TRU cores of the PGSFR are both inherently stable thanks to the temperature feedbacks.
Kurzfassung
Das koreanische Atomenergie-Forschungsinstitut (KAERI) hat einen 150 MWe-Prototyp eines Generation-IV-Natrium-gekühlten schnellen Reaktors (PGSFR) entwickelt. Das Designkonzept basiert hauptsächlich auf passiven Sicherheitsmechanismen. Derzeit ist es besonders wichtig die Dynamik und die Stabilität des Reaktors gegenüber kleinen Reaktivitätsstörungen während des Betriebs sicherzustellen. KAERI hat deshalb zur Stabilitätsanalyse des PGSFR den NuSTAB-Code entwickelt. Im NuSTAB-Code werden die Neutronen-kinetischen und thermo-hydraulischen Kopplungsgleichungen linearisiert um so die charakteristische Gleichung zu erhalten, die als verallgemeinertes Eigenwertproblem zur Bestimmung der Zerfallsrate, einem Indikator der Systemstabilität, behandelt wird. In diesem Beitrag wurde die Stabilität des PGSFR-Reaktors durch Anwendung punktkinetischer und räumlich-kinetischer Optionen im NuSTAB-Code analysiert. Die Systemantworten gegenüber Temperaturänderungen infolge des Dopplereffekts, thermischer Ausdehnung, Änderung der Kühlmitteldichte und das Gesamt-Feedback wurden untersucht. Die Ergebnisse zeigen, dass der anfängliche U-Kern und der endgültige TRU-Kern des PGSFR dank des Temperatur-Feedbacks inhärent stabil sind.
References
1 Brittan, R. O.: Some problems in the safety of fast reactor. ANL-5577, ANL, 195610.2172/4355722Suche in Google Scholar
2 Gialdi, E. et al.: Core stability in operating BWR: operational experience. Progress in Nuclear Energy15 (1985) 447–45910.1016/0149-1970(85)90070-8Suche in Google Scholar
3 US NRC: Information notice No. 88–39: Lasalle unit 2 loss of recirculation pumps with power oscillation event, June 15, 1988Suche in Google Scholar
4 Sandmeier, A. H.: The kinetics and stability of fast rectors with special considerations of nonlinearities. Doctorate thesis, Swiss Federal Institute of Technology in Zurich, 195910.2172/4236954Suche in Google Scholar
5 Depiante, E. V.: Stability analysis of a liquid-metal rector and its primary heat transport system. Nuclear Engineering and Design152 (1994) 361–37710.1016/0029-5493(94)90097-3Suche in Google Scholar
6 Bucys, K.; Svitra, D.: Modelling of nuclear reactors dynamics. Mathematical Modelling and Analysis4 (1999) 26–3210.3846/13926292.1999.9637107Suche in Google Scholar
7 Dokhane, A. et al.: Nonlinear stability analysis with novel BWR reduced order model. PHYSOR 2002, Seoul, Korea, October 7–10, 2002Suche in Google Scholar
8 Colombo, M. et al.: Transfer function modelling of the lead-cooled fast reactor (LFR) dynamics. Progress in Nuclear Energy52 (2010) 715–72910.1016/j.pnucene.2010.04.007Suche in Google Scholar
9 Akcasu, K. et al.: Mathematical methods in nuclear reactor dynamics. Academic Press, New York, 1971Suche in Google Scholar
10 Munoz-Cobo, J. L. et al.: Non-linear analysis of out of phase oscillations in boiling water reactors. Annals of Nuclear Energy23 (1996) 1301–133510.1016/0306-4549(96)00011-4Suche in Google Scholar
11 Karve, A. A.: Stability analysis of BWR nuclear-coupled thermal-hydraulics using a simple model. Nuclear Engineering and Design177 (1997) 155–17710.1016/S0029-5493(97)00192-1Suche in Google Scholar
12 US NRC: RELAP5/MOD3.3 code manuals. Nuclear Safety Analysis Division, NUREG/CR-5535, 2001Suche in Google Scholar
13 Joo, H. G. et al.: PARCS: A multi-dimensional two-group reactor kinetics code based on the nonlinear analytic nodal method. PU/NE-98-26, Purdue University, 1998Suche in Google Scholar
14 Costa, A. L. et al.: Simulation of an hypothetical out-of-phase instability case in boiling water reactor by RELAP5/PARCS coupled codes. Annals of Nuclear Energy35 (2008) 947–95710.1016/j.anucene.2007.08.019Suche in Google Scholar
15 US NRC: TRACE V5.0 Theory manual: Field equations, solution methods, and physical models. Division of Risk Assessment and Special Projects, 2007Suche in Google Scholar
16 Xu, Y. et al.: Application of TRACE/PARCS to BWR analysis. Annals of Nuclear Energy36 (2009) 317–32310.1016/j.anucene.2008.12.022Suche in Google Scholar
17 Hanggi, P.: Investigating BWR stability with a new linear frequency-domain method and detailed 3D neutronics. Ph.D. thesis, Swiss Federal Institute of Technology, 2001Suche in Google Scholar
18 Tsai, S. H. et al.: A critical eigenvalues tracing method for the small signal stability analysis of power systems. Energy and Power Engineering5 (2013) 677–68210.4236/epe.2013.54B131Suche in Google Scholar
19 Lindahl, S. O.: POLCA-Model description. ABB Report UR 86–192 Rev.4, 1986Suche in Google Scholar
20 Wulff, W. et al.: Description and assessment of RAMONA-3B Mod. 0 Cycle 4: a computer code with three dimensional neutron kinetics for BWR system transients. NUREG/CR-3664, 1984Suche in Google Scholar
21 Kang, C. M.: NuSTAB program manual. Internal Report, Korea Atomic Energy Research Institute, 2014Suche in Google Scholar
22 Chang, J.: Status of fast reactor technology development in Korea. The 45th IAEA TWG-FR Meeting, Beijing, China, June 20–22, 2012Suche in Google Scholar
23 Joo, H.: Status of fast reactor technology development in Korea. The 46th IAEA TWG-FR Meeting, Vienna, Austria, May 21–24, 2013Suche in Google Scholar
24 StaceyJr., W. M.: Space-time nuclear reactor kinetics. Academic Press, New York, 1960Suche in Google Scholar
25 Waltar, A. E. et al.: Fast spectrum reactors. Springer, 201110.1007/978-1-4419-9572-8Suche in Google Scholar
26 StaceyJr., W. M.: Variational flux synthesis methods for multigroup neutron diffusion theory. Nuclear Science and Engineering47 (1972) 449–46910.13182/NSE72-A22436Suche in Google Scholar
27 Yasinsky, J. B.; Kaplan, S.: Synthesis of three-dimensional flux shapes using discontinuous sets of trial functions. Nuclear Science and Engineering28 (1972) 426–437Suche in Google Scholar
28 Graham, J.: Fast reactor safety. Academic Press, New York and London, 1971Suche in Google Scholar
29 Tang, Y. S. et al.: Thermal analysis of liquid-metal fast breeder reactors. American Nuclear Society, 1978Suche in Google Scholar
30 Anderson, E. et al.: LAPACK user's guide. Third Edition, Society for Industrial and Applied Mathematics, 199910.1137/1.9780898719604Suche in Google Scholar
31 Kang, C. M.: NuSTAB verification and validation report. Internal Report, Korea Atomic Energy Research Institute, 2015Suche in Google Scholar
32 Lyapunov, A. M.: The general problem of the stability of motion. International Journal of Control55 (1992) 531–77310.1080/00207179208934253Suche in Google Scholar
33 Ha, P. N. V. et al.: A point dynamic model for stability analysis of the PGSFR. Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 07–08, 2015Suche in Google Scholar
34 Kim, J. D.: KAFAX-E66. Calculation Note No. NDL-23/01, Nuclear Data Evaluation Laboratory Internal Report, Korea Atomic Energy Research Institute, 2001Suche in Google Scholar
35 Macfarlane, R. E.: TRANSX-2: A code for interfacing MATXS cross section libraries to nuclear transport codes. LA-12312-MS, LANL, 1993Suche in Google Scholar
36 Alcouffe, R. E. et al.: User's guide for TWODANT: a code package for two-dimensional, diffusion-accelerated, neutron transport. LA-10049-M, LANL, 1990Suche in Google Scholar
37 Lim, J. Y.: Private communication. Korea Atomic Energy Research Institute, 2015Suche in Google Scholar
38 Kim, S. K. et al.: Thermal properties evaluation of U-Zr and U-Zr-Ce Alloys. Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 22, 2009Suche in Google Scholar
39 Golden, G. H.; Tokar, J. V.: Thermo physical properties of sodium. ANL-7323, ANL, August 196710.2172/4511962Suche in Google Scholar
© 2016, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Stability analysis of the Korean prototype Generation-IV sodium-cooled fast reactor using linear frequency domain approach
- Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors
- Steady state and transient analyses of MNSR reactor using RELAP5 code
- Protactinium-231 as a new fissionable material for nuclear reactors that can produce nuclear fuel with stable neutron-multiplying properties
- Assessment of pin-by-pin fission rate distribution within MOX/UO2 fuel assembly using MCNPX code
- Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models
- Experimental and numerical investigation on natural convection heat transfer in nanofluids
- Experimental studies in a single-phase parallel channel natural circulation system: preliminary results
- Calculation of PDS-XADS core closed-loop transfer function by using feedback with the lumped-model
- Calculation of nuclear reactivity using the generalised Adams-Bashforth-Moulton predictor corrector method
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Stability analysis of the Korean prototype Generation-IV sodium-cooled fast reactor using linear frequency domain approach
- Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors
- Steady state and transient analyses of MNSR reactor using RELAP5 code
- Protactinium-231 as a new fissionable material for nuclear reactors that can produce nuclear fuel with stable neutron-multiplying properties
- Assessment of pin-by-pin fission rate distribution within MOX/UO2 fuel assembly using MCNPX code
- Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models
- Experimental and numerical investigation on natural convection heat transfer in nanofluids
- Experimental studies in a single-phase parallel channel natural circulation system: preliminary results
- Calculation of PDS-XADS core closed-loop transfer function by using feedback with the lumped-model
- Calculation of nuclear reactivity using the generalised Adams-Bashforth-Moulton predictor corrector method