MNSR transient analysis using the RELAP5/Mod3.2 code

S. Dawahra; K. Khattab; F. Alhabit

Nuclear Engineering and Technology (Sep 2020)

MNSR transient analysis using the RELAP5/Mod3.2 code

S. Dawahra,
K. Khattab,
F. Alhabit

Affiliations

S. Dawahra: Corresponding author.; Nuclear Engineering Department, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria
K. Khattab: Nuclear Engineering Department, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria
F. Alhabit: Nuclear Engineering Department, Atomic Energy Commission, P.O. Box 6091, Damascus, Syria

Journal volume & issue: Vol. 52, no. 9
pp. 1990 – 1997

Abstract

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To support the safe operation of the Miniature Neutron Source Reactor (MNSR), a thermo-hydraulic transient model using the RELAP5/Mod3.2 code was simulated. The model was verified by comparing the results with the measured and the previously calculated data. The comparisons consisted of comparing the MNSR parameters under normal constant power operation and reactivity insertion transients. Reactivity Insertion Accident (RIA) for three different initial reactivity values of 3.6, 6.0, and 6.53 mk have been simulated. The calculated peaks of the reactor power, fuel, clad and coolant temperatures in hot channel were calculated in this model. The reactor power peaks were: 103 kW at 240 s, 174 kW at 160 s and 195 kW at 140 s, respectively. The fuel temperature reached its maximum value of 116 °C at 240 s, 124 °C at 160 s and 126 °C at 140 s respectively. These calculation results ensured the high inherently safety features of the MNSR under all phases of the RIAs.

Published in Nuclear Engineering and Technology

ISSN: 1738-5733 (Print)
Publisher: Elsevier
Country of publisher: Korea, Republic of
LCC subjects: Technology: Electrical engineering. Electronics. Nuclear engineering: Nuclear engineering. Atomic power
Website: http://www.journals.elsevier.com/nuclear-engineering-and-technology

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