Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors
-
A. Khedr
Abstract
In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.
Kurzfassung
Um die Bildung natürlicher Zirkulation in MTR Reaktoren nach Leistungsverlust besser zu verstehen, wurde ein experimenteller Prüfstand zur Simulation des Naturumlaufs gebaut. Der Prüfstand bestand aus zwei vertikal orientierten Zweigen, in einem der beiden wird der Kern simuliert durch zwei rechtwinklige, elektrisch beheizte, parallele Kanäle. Der andere Zweig simulierte den Teil des Rücklaufs, der an der Entstehung der natürlichen Zirkulation beteiligt ist. In der ersten Phase wurde eine Vielzahl experimenteller Durchläufe unter verschiedenen Bedingungen durchgeführt und die Messergebnisse und ihre Interpretation veröffentlicht. In dem jetzt vorliegenden Beitrag wurde das thermo-hydraulische Verhalten der experimentellen Anlage ergänzt durch eine theoretische Analyse mit Hilfe des RELAP5 Mod 3.3 Systemcodes. Die Analyse besteht aus zwei Teilen: im ersten Teil wird das RELAP5 Modell mit den gemessenen Werten validiert und im zweiten Teil werden die anderen, nicht gemessenen hydraulischen Parameters vorhergesagt und analysiert. Trotz des niedrigen Drucks des Prüfstands zeigen die Ergebnisse, dass RELAP5 das thermo-hydraulische Verhalten und das Begleitphänomen der Strömungsumkehr solcher Anlagen qualitativ vorhersagt. Quantitativ gibt es einen Unterschied zwischen vorhergesagten und gemessenen Werten, insbesondere bei der Oberflächentemperatur der Kanäle. Dieser Unterschied könnte auf die Unsicherheiten bei den anfänglichen Bedingungen der experimentellen Durchläufe, der Lage der Thermoelemente und auf das Wärmetransferpaket in RELAP5 zurückzuführen sein.
References
1 Alessandro, A.; D'Auria, F.: Thermal-Hydraulic System Codes in Nuclear Reactor Safety and Qualification Procedures. Science and Technology of Nuclear Installations, Volume 2008, ID 460795, 200810.1155/2008/460795Search in Google Scholar
2 Woodruff, W. L.; Hanan, N. A.; Smith, R. S.; Matos, J. E. A.: Comparison of the PARET/ANL and RELAP5/MOD3 codes for the analysis of IAEA Benchmarke transients. Proceeding of the International Meeting on Reduced Enrichment for Research and Test Reactors, pp. 1–11, Seoul, Republic of Korea, October 7–10, 1996Search in Google Scholar
3 Khedr, A.; Adorni, M.; D'Auria, F.: The effect of code user and boundary conditions on the RELAP calculations of MTR research reactors. Nuclear Technology & Radiation Protection1 (2005) 16–2210.2298/NTRP0501016KSearch in Google Scholar
4 Gee YangHan; KwiSeokHa: Modifications and Assessment of RELAP5/MOD3.2 for HANARO Thermal-Hydraulic Safety Analysis. Journal of the Korean Nuclear Society34 (2002) 455–467Search in Google Scholar
5 Antariksawan, A. R.; Zmitkova, J.; Huda, M. Q.; Allison, C. M.; Liu, T.; Hohorst, J. K.: Validation of RELAP/SCDAPSIM/MOD3.4 for research reactor applications. 13th International Conference on Nuclear Engineering, Beijing, China, May 16–20, 2005Search in Google Scholar
6 Adorni, M.; Bousbia-salah, A.; D'Auria, F.: Application of Best Estimate Thermalhydraulic Codes for the Safety Analysis of Research Reactors. International Conference of Nuclear Energy For New Europe, Portorož, Slovenia, Septemper 18–21, 2006Search in Google Scholar
7 Reis, P. A. L.; Costa, A. L.; Pereira, C.; Veloso, M. A. F.; Mesquita, A. Z.; Soares, H. V.; Barros, G. P.: Assessment of a RELAP5 model for the IPR- R1 TRIGA research reactor. Annals of Nuclear Energy37 (2010) 1341–135010.1016/j.anucene.2010.05.013Search in Google Scholar
8 Costa, A. L.; Reis, P. A. L.; Pereira, C.; Veloso, M. A. F.; Mesquita, A. Z.; Soares, H. V.: Thermal hydraulic analysis of the IPR-R1 research reactor using a RELAP5 model. Nuclear Engineering and Design240 (2010) 1487–149410.1016/j.nucengdes.2010.02.012Search in Google Scholar
9 Khedr, A.; Abdel-Hadi, E. A.; Talha, K. A.; Abdel-Latif, S. H.: Experimental Study of flow inversion in MTR upward flow research reactors. KERNTECHNIK79 (2014) 226–23610.3139/124.110425Search in Google Scholar
10 El-Din El-Morshedy, S.: Prediction, analysis and solution of the flow inversion phenomenon in a typical MTR-reactor with upward core cooling. Nuclear Engineering and Design241 (2011) 226–23510.1016/j.nucengdes.2010.10.006Search in Google Scholar
11 Khater, H. A.; Talha, K. A.; El-Fara, A. M.: Experimental Validation of a Thermal Hydraulic Model for ETRR-2 During Loss of Off Site Power. 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Pennsylvania, Sep. 30–Oct. 4, 2007Search in Google Scholar
© 2016, Carl Hanser Verlag, München
Articles in the same Issue
- 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
Articles in the same Issue
- 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
