Conceptual design station blackout and loss-of-flow accident analyses for the Advanced Neutron Source Reactor
- 1. Idaho National Engineering Lab., Idaho Falls, ID (United States)
Description
A system model of the Advanced Neutron Source Reactor (ANSR) has been developed and used to perform conceptual safety analyses. To better represent thermal-hydraulic behavior in the unique geometry and conditions of the ANSR core, three specific changes in the RELAP5/MOD3 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux correlation, and an interfacial drag correlation. The system model includes representations of the ANSR core, heat exchanger coolant loops, and the pressurizing and letdown systems. Analyses of ANSR station blackout and loss-of-flow accident scenarios are described. The results show that the core can survive without exceeding the flow excursion or critical heat flux thermal limits defined for the conceptual safety analysis, if the proper mitigation options are provided
Additional details
Publishing Information
- Journal Title
- Nuclear Technology
- Journal Volume
- 106
- Journal Issue
- 1
- Journal Page Range
- p. 31-45.
- ISSN
- 0029-5450
- CODEN
- NUTYBB
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 25058960
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- NEUTRON SOURCES; ORNL; R CODES; REACTOR ACCIDENTS; REACTOR SAFETY; REACTOR SIMULATORS; RESEARCH REACTORS; SAFETY ANALYSIS
- Descriptors DEC
- ACCIDENTS; ANALOG SYSTEMS; COMPUTER CODES; FUNCTIONAL MODELS; NATIONAL ORGANIZATIONS; PARTICLE SOURCES; RADIATION SOURCES; REACTORS; RESEARCH AND TEST REACTORS; SAFETY; SIMULATORS; US AEC; US DOE; US ERDA; US ORGANIZATIONS