Effect of reactor operator intervention during unprotected LOFA in a typical MTR
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S. M. A. Ibrahim
Abstract
The main objective for reactor engineering and inherent safety features is to keep the nuclear fuel in a thermally safe condition with enough safety margins during all operational states (normal-abnormal and accidental states). In this work, an accident analysis of the unprotected loss of flow transient in a typical material test reactor is presented. The effect of reactor operator intervention during the accident in case of manually triggering the shutdown system at different flow coast down ratios is studied. Flow trip scenarios are compared for flow trip ratios of 14, 15, 20, 30, 40, 50 and 60 %. The calculations were performed with the nuclear reactor analysis code PARET. The reactor stability region is described in terms of initial reactor power, core inlet temperature, and power peaking factor.
Kurzfassung
Das Hauptziel beim Betrieb eines Reaktors ist es, den Kernbrennstoff in allen Betriebszuständen in einem thermisch sicheren Zustand mit ausreichenden Sicherheitsmargen zu halten. In diesem Beitrag wird eine Unfallanalyse eines LOFA in einem typischen Materialprüfreaktor diskutiert. Dabei werden die Auswirkungen des Eingreifens des Reaktorpersonals während des Unfalls durch die manuelle Auslösung des Abschaltsystems bei unterschiedlichen Abströmverhältnissen untersucht. Dabei werden verschiedene Szenarien verglichen. Die Berechnungen wurden mit dem Analysecode PARET durchgeführt. Als Ergebnis wird der Reaktorstabilitätsbereich in Abhängigkeit von der anfänglichen Reaktorleistung, der Kerneintrittstemperatur und des Leistungsspitzenfaktors beschrieben.
References
1 Housiadas, C.: Simulation of loss-of-flow transients in research reactors. Annals of Nuclear Energy27 (2000) 1683–169310.1016/S0306-4549(00)00053-0Search in Google Scholar
2 Muhammad, F.: Simulation of uncontrolled loss of flow transients of a material test research reactor fuelled with low and high enriched uranium dispersion fuels. Annals of Nuclear Energy37 (2010) 582–59110.1016/j.anucene.2009.12.019Search in Google Scholar
3 Kazeminejad, H.: Thermal-hydraulic modeling of flow inversion in a research reactor. Annals of Nuclear Energy35 (2008) 1813–181910.1016/j.anucene.2008.05.006Search in Google Scholar
4 Salama, A.; El-Morshedy, S. E.-D.: CFD analysis of flow blockage in MTR coolant channel under loss-of-flow transient: Hot channel scenario. Progress in Nuclear Energy55 (2012) 78–9210.1016/j.pnucene.2011.11.005Search in Google Scholar
5 El-Morshedy, S. E.-D.: 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
6 El-Morshedy, S. E.-D.: Thermal-hydraulic modeling and analysis of a tank in pool reactor for normal operation and loss of flow transient. Progress in Nuclear Energy61 (2012) 78–8710.1016/j.pnucene.2012.07.005Search in Google Scholar
7 Khater, H.; El-Morshdy, S. E.-D.; Abdelmaksoud, A.: Simulation of unprotected LOFA in MTR reactors using a mix CFD and one-d computation tool. Annals of Nuclear Energy83 (2015) 376–38510.1016/j.anucene.2015.03.041Search in Google Scholar
8 Obenchain, C. F.: PARET – A Program for the Analysis of Reactor Transients. ACE Research and Development Report, IDO-1728, 196910.2172/4812807Search in Google Scholar
9 ScottJr, R.; Hale, C.; Hagen, R.: Transient tests of fully enriched uranium oxide stainless steel plate type c-core in the SPERT-III Reactor. Data Summary Report, IDO-17223, (1967)Search in Google Scholar
10 Woodruff, W. L.: A kinetics and thermal hydraulics capability for the analysis for research reactor. ANL, (1983)Search in Google Scholar
11 Khater, H.; Abu-EL-Maty, T.; El-Morshdy, S. E.-D.: Thermal-hydraulic modeling of reactivity accident in MTR reactors. Annals of Nuclear Energy34 (2007) 732–74210.1016/j.anucene.2007.03.012Search in Google Scholar
12 Lamarsh, J. R.: Introduction to Nuclear Reactor Theory. Addison-Wesley, Reading, MA, (1966)Search in Google Scholar
© 2019, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- New research reactor protection system
- Ultimate response guideline strategy evaluation for Maanshan power plant
- Application of two alternative shutdown severe accident management guideline (SSAMG) entry conditions for CPR1000
- Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code
- Numerical study on cut-off diameter of aerosol particle for filtered containment venting system in nuclear power plant
- Reanalysis of environmental qualification radiation parameters for Kuosheng nuclear power plant
- Effect of reactor operator intervention during unprotected LOFA in a typical MTR
Articles in the same Issue
- Contents/Inhalt
- Contents
- Technical Contributions/Fachbeiträge
- New research reactor protection system
- Ultimate response guideline strategy evaluation for Maanshan power plant
- Application of two alternative shutdown severe accident management guideline (SSAMG) entry conditions for CPR1000
- Simulation and analysis of the Loss of Flow Accident (LOFA) scenarios for an open pool type research reactor by using the RELAP5/MOD3.2 code
- Numerical study on cut-off diameter of aerosol particle for filtered containment venting system in nuclear power plant
- Reanalysis of environmental qualification radiation parameters for Kuosheng nuclear power plant
- Effect of reactor operator intervention during unprotected LOFA in a typical MTR
