SAS4A/SASSYS-1 (Reactor Dynamics and Safety Analysis Code)
Table of Contents
The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs). With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in advanced reactor safety analysis capabilities for the U.S. DOE.
SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. Although SAS4A and SASSYS-1 are generally portrayed as two computer codes, they have always shared a common code architecture, the same data management strategy, and the same core channel representation. Subsequently, the two code branches were merged into a single code referred to as SAS4A/SASSYS-1.
"Protected Loss of Flow Transient Simulation" video available.
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A version of SAS4A/SASSYS-1 with a reduced feature-set, Mini SAS, is freely available for academic and non-commercial use. Mini SAS is built from the same source as SAS4A/SASSYS-1, but excludes severe accident and steam plant models. It is also limited to five core channels, which is adequate for most analyses.
SAS4A/SASSYS-1 has been coupled to a variety of analysis and optimization tools, such as Star-CCM+, Dakota, RAVEN, SAM (the System Analysis Module under development as part of the U.S. DOE NEAMS program), and PDC (the Argonne Plant Dynamics Code that models S-CO2 Brayton cycles). Several benchmark models have been developed for validation of whole-plant passive safety response based on EBR-II tests conducted in the 1980s. Two of these tests, Shutdown Heat Removal Tests 17 and 45R, are the basis of an IAEA CRP led by Argonne. DOE and Argonne are currently preparing additional validation models based on FFTF testing conducted in 1986. Publically available literature on these efforts are listed below, in Relevant Publications.
The code is currently supported on macOS, Windows, and Linux platforms with the x86 architecture. Source code is compliant with Fortran 2003 standards and can be compiled with any standards-compliant Fortran compiler.
The code manual and details on the latest code updates can be found on the SAS4A/SASSYS-1 public wiki.
To obtain a license for SAS4A/SASSYS-1, prospective users need to contact the Technology Development and Commercialization Division of Argonne. Licenses may be obtained for executable only or for full source code access. Once a fully executed license has been approved, the code manager handles distribution to the licensee. Licenses for Mini SAS (executable copies only) are also available for non-commercial use under a general license agreement and can be obtained directly from the code manager.
- Single-pin channel models for rapid evaluation of transients
- Detailed thermal-hydraulic sub-channel models for subassembly pin bundles
- Support for three-dimensional visualization of sub-channel temperatures
- Support for liquid-metal coolants such as sodium, NaK, lead and LBE, as well as other single-phase coolants
- Full-plant coolant system models to simulate passive heat removal and natural shutdown
- Oxide fuel models for fuel melting, in-pin motion, pin failure, and ex-pin fuel dispersal and freezing
- Metal fuel models for fuel-clad eutectic formation and cladding failure
- Reactor point kinetics with comprehensive treatment of reactivity feedback effects as per first-order perturbation theory
- High-fidelity decay heat models
- Built-in support for ANS standard decay heat properties
- Built-in support for alternative coolants in decay heat removal loops
- Support for line-based comments in input files
- Support for an unlimited number of time steps
- Support for coupling to third-party computational fluid dynamics tools (such as STAR-CCM+) for representing thermal stratification in large volumes
- Support for coupling to DIF3D-K for reactor spatial kinetics
- Detailed reactor and plant control systems
Development of the SAS series of codes began in the mid-1960s to model the initiating phases of hypothetical core disruptive accidents. SAS1A originated from a sodium-boiling model and included single- and two-phase coolant flow dynamics, fuel and cladding thermal expansion and deformation, molten fuel dynamics, and a point kinetics model with reactivity feedback. By 1974, SAS evolved to the SAS2A computer code which included a detailed multiple slug and bubble coolant boiling model which greatly enhanced the ability to simulate the initiating phases of loss of flow (LOF) and transient overpower (TOP) accidents up to the point of cladding failure and fuel and cladding melting.
The SAS3A code added mechanistic models of fuel and cladding melting and relocation. This version of the code was used extensively for analysis of accidents in the licensing of the Fast Flux Test Facility. In anticipation of loss of flow and transient overpower analysis requirements for licensing of the Clinch River Breeder Reactor Plant, new fuel element deformation, disruption, and material relocation models were written for the SAS4A version of the code which saw extensive validation against TREAT M-Series test data. In addition, a variant of SAS4A named SASSYS-1 was developed with the capability to model ex-reactor coolant systems to permit the analysis of accident sequences involving or initiated by loss of heat removal or other coolant system events. This allows the simulation of whole-plant dynamics feedback for both shutdown and off-normal conditions.
Although SAS4A and SASSYS-1 are sometimes portrayed as two computer codes, they have always shared common code architectures, the same data management strategy, and the same core channel representation. Subsequently, the two code branches were merged into a single code referred to as SAS4A/SASSYS-1. Version 2 of the SAS4A/SASSYS-1 code was distributed to Germany, France, and Japan in the late 1980s, and it serves as a common tool for international oxide fuel model developments.
Beyond the release of SAS4A/SASSYS-1 v 2.1, revisions to SAS4A/SASSYS-1 continued throughout the Integral Fast Reactor program between 1984 and 1994 culminating with the completion of SAS4A/SASSYS-1 v 3.0 in 1994. During this time, the modeling emphasis shifted towards metallic fuel and accident prevention by means of inherent safety mechanisms. This resulted in 1) addition of new models and modification of existing models to treat metallic fuel, its properties, behavior, and accident phenomena, and 2) addition and validation of new capabilities for calculating whole-plant design basis transients, with emphasis on the EBR-II reactor and plant, the IFR prototype. The whole-plant dynamics capability of the SASSYS-1 component plays a vital role in predicting passive safety feedback. Without it, meaningful boundary conditions for the core channel models are not available, and accident progression is not reliably predicted.
By the mid 1990s, SAS4A/SASSYS-1 v 3.1 had been completed as a significant maintenance update, but it was not released until 2012. In the time since the development of Version 3, a variety of modeling additions and enhancements have been made to meet U.S. DOE programmatic needs. Additional information on the current version (Version 5) and latest updates to the code can be found on the SAS4A/SASSYS-1 public wiki.
- G. Zhang, et al., “Dakota-SAS4A/SASSYS-1 Coupling for Uncertainty Quantification and Optimization Analysis,” in Proceedings of M&C 2017 – International Conference on Mathematics & Computational Methods Applied to Nuclear Science & Engineering, Jeju, Korea, 2017.
- T.H. Fanning and R. Hu, "Coupling the System Analysis Module with SAS4A/SASSYS-1", ANL/NE-16/22, ANL-ART-74, Argonne National Laboratory, September 2016.
- A.J. Brunett and T.H. Fanning, “Uncertainty Quantification in Advanced Reactors: The Coupling of SAS4A/SASSYS-1 with RAVEN and Dakota,” in Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP 2016), San Francisco, CA, 2016.
- T. Sumner and A. Moisseytsev, “Simulations of the EBR-II Tests SHRT-17 and SHRT-45R,” in Proceedings of the 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16), Chicago, IL, 2015.
- A. Moisseytsev and J. Sienicki, “Dynamic Analysis of SCO2 Cycle Control with Coupled PDC-SAS4A/SASSYS-1 Codes,” in Proceedings of the 20th International Conference on Nuclear Engineering (ICONE20), Anaheim, CA, 2012.
- J. W. Thomas, et al., “Validation of the Integration of CFD and SAS4A/SASSYS-1: Analysis of EBR-II Shutdown Heat Removal Test 17,” in Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP 2012), Chicago, IL, 2012.
Protected Loss of Flow Transient Simulation
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Watch this video on YouTube: Protected Loss of Flow Transient Simulation
Tests carried out by Argonne were used to perform validation for advanced safety simulations, such as the one shown here from SAS4A/SASSYS-1. This image shows the calculated fuel, cladding, coolant, and structure temperatures for the XX09 experimental assembly and its six neighbors during a full-power loss of flow accident.
Last Modified: Tue, April 4, 2017 9:50 AM