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Modeling the Accumulators of the Advanced Neutron Source Reactor
Abstract
An improved capability for modeling the transient behavior of the unique Advanced Neutron Source Reactor (ANSR) accumulator design has been developed. The ANSR accumulator behavior may now be simulated using an external model that is coupled to a RELAP5/MOD3 ANSR facility system model via a parallel virtual machine (PVM) connection. Existing RELAP5 code models were found to provide marginal results for simulating behavior of the unique ANSR accumulator design, and an improved accumulator modeling capability was desired. A new model for representing the ANSR accumulators, assessment and demonstration of that model, and the general methods by which external models may be coupled with RELAP5 system models using PVM connections are described.
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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.
A RELAP5 system model for the Advanced Neutron Source Reactor has been developed for performing conceptual safety analysis report calculations. To better represent thermal-hydraulic behavior of the core, three specific changes in the RELAP5 computer code were implemented: a turbulent forced-convection heat transfer correlation, a critical heat flux (CHF) correlation, and an interfacial drag correlation. The model consists of the core region, the heat exchanger loop region, and the pressurizing/letdown system region. Results for three loss-of-coolant accident analyses are presented: (1) an instantaneous double-ended guillotine (DEG) core outlet break with a cavitating venturi installed downstream of the core, (b) a core pressure boundary tube outer wall rupture, and (c) a DEG core inlet break with a finite break-formation time. The results show that the core can survive without exceeding the flow excursion of CHF thermal limits at a 95% probability level if the proper mitigation options are provided.
The ongoing preconceptual and conceptual reactor design of the Advanced Neutron Source (ANS) is explored. The ANS is being designed for materials sciences, isotope production, and fundamental physics research. A reactor design based on previously developed technology can meet the performance requirements set by the user community for a new ANS to serve all fields of neutron science. These requirements include the capability of producing a peak thermal neutron flux over five times higher than that in use at any currently operating steady-state facility. Achievement of these ultrahigh flux levels involves many interesting aspects of reactor design.
The reactor characteristics of the current preconceptual reference design are presented. The attainment of this design was reached by following a design strategy that best met the safety and user requirements. The design has evolved over the last 5 yr from two concepts proposed in 1985. The trade-offs and selection of many reactor parameters are described to illustrate how and why the current design was achieved. Further reactor design is planned, leading to an ANS operating by 1999 for use by scientists of many disciplines.
Software Evaluation and Validation Report of RELAP5/NPR Version 0 (U)
- W E Bickford
- K F Chen
- W L Howarth
- Z H Qureshi
PVM 3.0 User’s Guide and Reference Manual”, ORNL/TM-12187
- A Geist
- A Begelin
- J Dongarra
- W Jiang
- R Manchek
- V Sunderam
Multiprocessor Based Simulation of Degraded Core and Containment Responses
- K A Smith
RELAP5/CONTAIN Coupling”, presented at RELAP5 International Users Seminar
- R P Martin