This paper outlines a system-level safety analysis procedure for research reactors incorporating sensitivity and uncertainty components. The loss of flow accident is selected as a typical design basis accident to demonstrate the analysis procedure.  The conceptual NIST (National Institute of Standards and Technology) horizontally split-core based research reactor is used a typical research reactor example in the study. Two system level dynamics codes, Relap5-3D and PARET, are employed in this work in a comparison study manner. The primary objective of the present work is to demonstrate the analysis capability of integrating sensitivity and uncertainty information in addition to traditional predictions of the RELAP5-3D models for the study of the thermal-hydraulics (T/H) safety characteristics of research reactors under accidental transient scenarios.

In this paper, a nodalization of the reactor core with natural circulation loop of the conceptual NIST reactor was developed by RELAP5-3D. The protected fast and slow loss of flow (LOF) accidents have been modeled and investigated. The coolant was flowing in a downward direction during forced convective cooling in normal operating conditions, while a reversed natural circulation of coolant was established during the LOF accidents and controlled by buoyancy forces. Key response parameters including the peak cladding and peak coolant temperatures were obtained and compared. The preliminary research results showed the transient predictions yielded by the two system codes mentioned above have a good agreement for the analysis of the loss of flow accidents. However, attributed to the mechanism modeling differences in the establishment of natural circulation, the results have shown some discrepancies in the flow and temperature estimations during the flow reversal period in the course of stabilizing natural flow circulation. To better understand the discrepancies observed in the simulations, sensitivity and uncertainty analyses were performed by coupling the Relap5-3D model and RAVEN, a probabilistic risk assessment code developed by INL. The sensitivity analysis reveals the significances of key figure of merits such as the peak cladding temperature varies with different boundary and initial parameters in both normal operation and design basis transients. The uncertainty analysis informs the deviations of the responses contributed by the errors of various input components. Both the sensitivity and uncertainty information will be incorporated into a safety analysis framework as part of the safety characteristic predictions produced by the framework.


Sensitivity and Uncertainty Information Incorporated Loss of Flow Accident Analyses for Research Reactors

Division / Sponsors: 
Power Division & Nuclear Engineering Division (NED)

Power Conference
Paper Proceedings: https://asmedigitalcollection.asme.org/POWER/POWER2020/volume/83747

International Conference on Nuclear Engineering
Paper Proceedings: https://asmedigitalcollection.asme.org/ICONE/ICONE2020/volume/83761

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