Steady state thermal hydraulic calculations for MTR plate-type research reactors
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I. D. Abdelrazek
Abstract
A computation code THERMOHYD has been developed to perform the steady state thermal-hydraulic calculations of the MTR type research reactors fueled with plate-type-fuel, in the ranges of pressures and temperatures typical for MTR reactors. The code calculates the coolant and clad surface temperatures along the cooling channel, as well as the critical heat flux and provides values for the critical phenomena parameters (RDR, ONBR and DNBR ratios) for reactor cores. The code takes into account the variation of the physical properties of the coolant with temperature. It also accounts for the uncertainties due to fuel fabrication tolerances, possible deviations in the operational measurements, the errors in the experimental correlations used in the analysis, and the errors due to simplifications made in the calculations which have been considered in the analysis. The THERMOHYD code was tested against the well known PARET code and proved to be accurate and more conservative in some aspects.
Kurzfassung
Ein Coputercode THERMOHYD wurde entwickelt, zur Berechnung der Thermohydraulik im stationären Zustand für MTR Forschungsreaktoren vom Plattentyp bei den für diese Reaktoren üblichen Drucken und Temperaturen. Der Code berechnet die Temperaturen des Kühlmittels und die Temperatur an der Oberfläche der Brennelementhülle entlang des Kühlkanals ebenso wie den kritischen Wärmefluss und liefert Werte für die Parameter der kritischen Phänomene (RDR, ONBR und DNBR Verhältnisse) des Reaktorkerns. Der Code berücksichtigt die Schwankung der physikalischen Eigenschaften des Kühlmittels mit der Temperatur. Er berücksichtigt auch die Unsicherheiten aufgrund der Herstellungstoleranzen des Brennstoffs, mögliche Abweichungen bei den Routinemessungen, die Fehler bei den in der Analyse verwendeten experimentellen Korrelationen sowie die Fehler aufgrund von Vereinfachungen, die bei den Berechnungen gemacht wurden und die bei der Analyse berücksichtigt werden müssen. THERMOHYD Code wurde mit Hilfe des bekannten PARET Codes getestet und hat sich als genau und in Bezug auf einige Aspekte als eher konservativ herausgestellt.
References
1 AEA/INVAPS.E.: Standard Fuel Assembly. ETRR-2 Document: 0767-0710-3AAIT-963-1A, Egypt, 1995Suche in Google Scholar
2 Nuclear Power Engineering, M. M. El-Wakil, McGraw Hill, 1962Suche in Google Scholar
3 Abdelrazek, I.; Zidan, W.: A program for the Thermo-hydraulic Calculations of MTR Research Reactors. AEA Report 481, 2008Suche in Google Scholar
4 Principles of Unit Operations, A. S.Foust, Ed., J. Wiley & Sons, 1964Suche in Google Scholar
5 IAEA-TECDOC-233, Research Reactor Core Conversion from the use of Highly Enriched Uranium to the use of Low Enriched Uranium, Appendix D, Generic Enrichment Reduction Calculations, (France), 1980Suche in Google Scholar
6 Principles of Heat Transfer, 2nd Ed, F. Kreith, International Text Book Co., 1967Suche in Google Scholar
7 INVAPS.E., TERMIC 1 H v 3.1: A Program for the Calculus and Thermo-hydraulic design of Reactor Cores, MTR_PC package, Argentina (1995)Suche in Google Scholar
8 Fabrega, J.: Le Calcul Termique des Reacteurs de Recherche Refroidis Par Eau. Rapport CEA R-4114, France, 1971Suche in Google Scholar
9 Review of Thermal-Hydraulic Methodology and design Margins, CAB-CNEA, S. C. de Bariloche, ArgentinaSuche in Google Scholar
10 Hergenrede, D.; Villarino, E.; Durione, C.: Analysis of the Second Refueling. ETRR-2, Atomic Energy Authority of Egypt, Report 0767-0740-3TANU-193-1O 2000, 2000Suche in Google Scholar
11 Abdelrazek, I.et al.: Steady State Thermal-hydraulic analysis for ETRR-2 Research Reactor. Kerntechnik 69 (2004) 3Suche in Google Scholar
12 Obenchain, C. F.: PARET-A Program for the Analysis of Reactor Transients, ACE Research and Development Report, IDO-17282, 1969Suche in Google Scholar
© 2010, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Adaptation of the FEMAXI-6 code and RBMK fuel rods model testing employing the best estimate approach
- In-reactor measurements of fuel centerline temperature variation during power change
- Pertinence of tests used for conformity assessments of fire resistance of nuclear power plant components – opinion of several institutions in Europe
- Analysis of the instability of OTSG of movable nuclear power devices by using the multivariable frequency domain method
- Steady state thermal hydraulic calculations for MTR plate-type research reactors
- Production cross sections of medical 110,111In radionuclides
- Nuclear level density parameters of some largely deformed nuclei of medical radionuclides in low-lying collective excitation modes
- A discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium
- Kritikalitätssicherheit der Endlagerung ausgedienter Kernbrennstoffe
- Technical Notes/Technische Mitteilungen
- Methods for thin foil thickness determination by using alpha spectroscopy
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- Adaptation of the FEMAXI-6 code and RBMK fuel rods model testing employing the best estimate approach
- In-reactor measurements of fuel centerline temperature variation during power change
- Pertinence of tests used for conformity assessments of fire resistance of nuclear power plant components – opinion of several institutions in Europe
- Analysis of the instability of OTSG of movable nuclear power devices by using the multivariable frequency domain method
- Steady state thermal hydraulic calculations for MTR plate-type research reactors
- Production cross sections of medical 110,111In radionuclides
- Nuclear level density parameters of some largely deformed nuclei of medical radionuclides in low-lying collective excitation modes
- A discrete ordinates operator for the diffuse transmission of a mixed neutron beam through a multilayered medium
- Kritikalitätssicherheit der Endlagerung ausgedienter Kernbrennstoffe
- Technical Notes/Technische Mitteilungen
- Methods for thin foil thickness determination by using alpha spectroscopy
