Lumped parameters analysis of the IAEA research reactor benchmark problem
-
M. A. Gaheen
Abstract
A simple model has been developed to predict the transient behaviour of research reactors with simple means. The developed model uses a lumped parameters approach for the kinetics and heat transfer modeling with continuous feedback reactivities. The model is used for the analysis of the International Atomic Energy Agency (IAEA) 10MW MTR research reactor benchmark problem. Transient responses to reactivity insertion and loss of coolant flow are presented and analyzed. The model predictions are accurate enough compared with the calculations conducted in various institutions using different codes. It is shown that the model can provide accurate predictions as long as the clad temperature does not exceed the ONB (Onset of Nucleate Boiling) temperature. However, the results are very encouraging and the model is useful for practical purposes.
Kurzfassung
Ein einfaches Modell zur Vorhersage des Transientenverhaltens von Forschungsreaktoren wurde entwickelt. Das Modell verwendet einen Lumped-Parameter-Ansatz zur Modellierung von Kinetik und Wärmetransfer mit kontinuierlichen Feedback Reaktivitäten. Das Modell wird verwendet zur Analyse des IAEA Benchmark-Problems bei 10MW MTR Forschungsreaktoren. Das Transientenverhalten gegenüber Reaktivitätseintrag und Verlust des Kühlmittels wird analysiert und dargestellt. Die Modellvorhersagen sind hinreichend genau, verglichen mit Berechnungen anderer Institutionen bei Verwendung unterschiedlicher Codes. Es wird gezeigt, dass das Modell genaue Vorhersagen liefert solange die Temperatur unterhalb der jetzigen liegt, bei der das Blasensieden beginnt.
References
1Obenchain, C. F.: PARET-A program for the analysis of reactor transients. Idaho National Engineering Laboratory, Report IDO-17282, 1969Suche in Google Scholar
2Bousbia-Salah, A.; Hamidouche, T.: Analysis of the IAEA research reactor benchmark problem by the RETRAC code. Nuclear Engineering and Design235 (2005) 661–674Suche in Google Scholar
3IAEA-TECDOC-643: Research Reactor Core Conversion Guidebook. Vol. 3: Analytical Verification, 1992Suche in Google Scholar
4McGurire, M. J.: An Analysis of the Proposed MITR-III Core to Establish Thermal-Hydraulic Limits at 10 MW. Ph. D. theses MIT, 1995Suche in Google Scholar
5Housiadas, C.: Lumped parameters analysis of coupled kinetics and thermal-hydraulics for small reactors. Annals of Nuclear Energy, 29, 1315–1325, 200210.1016/S0306-4549(01)00107-4Suche in Google Scholar
6Nasir, R.; Mirza, N. M.; Mirza, S. M.: Sensitivity of reactivity insertion limits with respect to safety parameters in a typical MTR. Annals of Nuclear Energy26 (1999) 1517–1535Suche in Google Scholar
7Woodruff, 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 benchmark transients. International Meeting on Reduced Enrichment for Research and Test Reactors, Seoul, South Korea, October 7–10, 1996Suche in Google Scholar
© 2006, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- The multipurpose thermalhydraulic test facility TOPFLOW: an overview on experimental capabilities, instrumentation and results
- Uncertainty in cross-section calculations for reactions induced by neutrons with energy above 0.1 MeV
- Determination of the exposure build-up factor in a slab using the LTSN method
- Lumped parameters analysis of the IAEA research reactor benchmark problem
- Assessment of the look-up table using the tubular and bundle CHF data and modification of the bundle correction factor
- A finite element model for static strength analysis of CANDU fuel bundle
- Effect of packing fraction variations on the multiplication factor in pebble-bed nuclear reactors
- Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects
- Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina Nuclear Power Plant
- Analyses of severe accident scenarios in RBMK-1500
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- The multipurpose thermalhydraulic test facility TOPFLOW: an overview on experimental capabilities, instrumentation and results
- Uncertainty in cross-section calculations for reactions induced by neutrons with energy above 0.1 MeV
- Determination of the exposure build-up factor in a slab using the LTSN method
- Lumped parameters analysis of the IAEA research reactor benchmark problem
- Assessment of the look-up table using the tubular and bundle CHF data and modification of the bundle correction factor
- A finite element model for static strength analysis of CANDU fuel bundle
- Effect of packing fraction variations on the multiplication factor in pebble-bed nuclear reactors
- Design of a dry cask storage system for spent LWR fuels: radiation protection, subcriticality, and heat removal aspects
- Radiological and thermal characteristics of CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks for spent nuclear fuel storage at Ignalina Nuclear Power Plant
- Analyses of severe accident scenarios in RBMK-1500
