Software

SAS4A (Reactor Dynamics and Safety Analysis Codes)

Standard Code Description

1. Name of Program:
   SAS4A
2. Computer for Which Program is Designed and Other Machine Version Packages Available:
   Mainframe (IBM, CRAY, CDC, etc.), Unix Workstation (Sun, IBM RISC, HP, SG), or Personal Computer (IBM PC) with FORTRAN Compiler.
   
3. Description of Problem Solved:
   SAS4A is designed to perform deterministic analysis of severe accidents in liquid metal cooled reactors (LMRs). Detailed, mechanistic models of steady-state and transient thermal, hydraulic, neutronic, and mechanical phenomena are employed to describe the response of the reactor core and its coolant, fuel elements, and structural members to accident conditions caused by loss of coolant flow, loss of heat rejection, or reactivity insertion. The initiating phase of the accident is modeled, including coolant heating and boiling, fuel cladding failure, and fuel melting and relocation. SAS4A analysis is terminated upon loss of subassembly hexcan integrity. The objective of SAS4A analysis is to quantify severe accident consequences as measured by the generation of energetics sufficient to challenge reactor vessel integrity, leading possibly to public health and safety risk. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models in SAS4A were subsequently extended and specialized to metallic fuel clad with advanced alloys.
   
4. Method of Solution:
   In space, each SAS4A channel represents one or more subassemblies with either a single pin model or a multiple pin model. Many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior model, integrated with fission product production, release, and transport models, provides prediction of fuel element dimensional changes and cladding failure. Fuel and cladding melting and subsequent relocation are described with multiple-component fluid dynamics models, with material motions driven by pressures from coolant vaporization, fission gas liberation, and fuel and cladding vaporization. Reactivity feedbacks from fuel heating (axial expansion and Doppler), coolant heating and boiling, and fuel and cladding relocation are tracked with first order perturbation theory. Reactivity effects from reactor structural temperature changes yielding radial core expansion are modeled. Changes in reactor power level are computed with point kinetics. Numerical models used in the code modules range from semi-implicit to explicit. The coupling of modules in time is semi-explicit within a multiple-level time step framework.
5. Restrictions on the Complexity of the Problems:
   In any channel, there are maximums of 24 axial heat transfer nodes in the core and axial blankets and 49 axial coolant hydraulics nodes. The number of channels is limited only by the size of the computer memory.
6. Typical Running Time:
   Running times depend on the complexity of the model and the physical phenomena being analyzed. A few-channel reactor model using only pin heat transfer, single phase coolant dynamics, and reactor point kinetics physical models will generally run orders of magnitude faster than real time on modern computing hardware. A many-channel model using two-phase coolant dynamics and fuel melting and relocation physical models take significantly longer, with running times that depend on problem complexity.
7. Unusual Features of the Program:
   The physical models in SAS4A are highly detailed numerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.
8. Related and Auxiliary Programs:
   Much of the reactor core and coolant loop thermal hydraulic models in SAS4A are shared with the SASSYS-1 computer code.
9. Status:
   SAS4A Version 3.1 is available for production use at Argonne National Laboratory in the Nuclear Engineering Division. Earlier versions have been exported to domestic U.S. DOE contractors and to research organizations in foreign countries. The SAS4A/SASSYS-1 code package continues to undergo development in response to advanced fast reactor simulation needs.
10. References:
    1. J. E. Cahalan et al., "Advanced LMR Safety Analysis Capabilities in the SASSYS-1 and SAS4A Computer Codes," Proceedings of the International Topical Meeting on Advanced Reactors Safety, Pittsburgh, PA, April 17­21, American Nuclear Society, 1994.
    2. J. E. Cahalan and T. Wei, "Modeling Developments for the SAS4A and SASSYS Computer Codes," Proceedings of the International Fast Reactor Safety Meeting, Snowbird, UT, August 12­16, American Nuclear Society, 1990.
11. Machine Requirements:
    The length of the combined SAS4A/SASSYS-1 executable on the Sun Unix system is about 7.2 Mbytes, and a data buffer of about 200 Kbytes for each channel is required. Disk storage for potentially large ASCII print and binary plotting data storage files is required.
12. Programming Languages Used:
    Standard FORTRAN 77 is used. System dependent routines may be supplied for dynamic memory allocation, timing, and system and user identification.
13. Operating System:
    No special requirements other than a FORTRAN compiler and the usual linker/loader facilities.
14. Other Programming or Operating Information or Restrictions:
    The distribution of the SAS4A computer code and its documentation are subject to U.S. DOE Applied Technology regulations.
15. Name and Establishment of Author or Contributor:
    • J. E. Cahalan
    • Nuclear Engineering Division
      Argonne National Laboratory
      9700 South Cass Avenue
      Argonne, Illinois 60439
16. Materials Available:
    • FORTRAN Source Code
    • Example Problems Input Data and Printed Output
    • Five Volume Technical Report for Version 3.0 containing detailed model descriptions and user guide.
17. Sponsor:
    U.S. Department of Energy, Office of Nuclear Energy, Science, and Technology.

Last Modified: Wed, September 15, 2010 7:05 PM