United States Nuclear Regulatory Commission - Protecting People and the Environment

Research Reactor ‘MARIA’ Primary Cooling Loop Transient Analysis Using RELAP5 Mod 3.3 (NUREG/IA-0443)

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Publication Information

Manuscript Completed: February 2014
Date Published: June 2014

Prepared by:
M. Dabrowski, E. Staron

Panstwowa Agencja Atomistyki
(National Atomic Energy Agency)
Krucza 36
00-522 Warszawa, POLAND

K. Tien, NRC Project Manager

Division of Systems Analysis
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by:
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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Abstract

Although the Research Reactor ‘Maria’ is under operation for a considerable amount of time, the safety analysis of the reactor had been performed using a specialized thermal-hydraulic code adapted to the RR ‘Maria’ requirements only, until the Polish regulatory body – PAA joined CAMP and received access to RELAP and TRACE. This possibility coincided with modernization works in the primary cooling loop of the research reactor that resulted as a consequence of the Global Threat Reduction Initiative (GTRI) which obliged the operator of the research reactor to change the fuel from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU). The first version of the analysis was limited to only one fuel channel and was presented in NUREG/IA-0422 published in 2013. A continuation of those works is presented in this report covering an analysis of the whole primary cooling circuit of the research reactor. The steady state results were assessed as satisfactory. The analysis included a number of cases covering a Small Break, Medium Break, Large Break and Double Ended Guillotine Break Loss of Coolant Accident. The results received are presented on graphs and proved that in most cases no danger to fuel integrity exists. However, in case of Medium Break LOCA a fuel degradation risk has been envisaged. This scenario will be analysed more thouroughly in the future. Loss of Fluid and Loss of Heat Sink were also analysed – in these transients no risk to fuel integrity was found. A sensitivity analysis is also included.

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