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SAS4A/SASSYS-1 (Reactor Dynamics and Safety Analysis Codes)

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SAS4A | SASSYS-1

Standard Code Description for SAS4A

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  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 17-21, 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 12-16, 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.

Standard Code Description for SASSYS-1

  1. Name of Program:
    SASSYS-1
  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:
    SASSYS-1 is designed to perform deterministic analysis of design basis and beyond-design basis accidents in liquid metal cooled reactor (LMR) plants. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, neutronic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by loss of coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents are modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loops performance including natural circulation, and decay heat removal. SASSYS-1 analysis is terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of SASSYS-1 analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models in SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys.
  4. Method of Solution:
    In space, each SASSYS-1 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 the margin to cladding failure. Thermal-hydraulic models of the reactor and intermediate coolant loops analyze heat removal from both forced and natural circulation, with transient performance of loop components including pumps, heat exchangers, valves, and plena. The balance-of-plant thermal-hydraulic model performs transient simulation of the feedwater/steam system. Both the sodium loops model and the balance-of-plant model are integrated with the plant protection and control system model, which is used to simulate the performance of reactor scram systems, and controllers on pumps, valves, and decay heat removal systems. Reactivity feedbacks from fuel heating (axial expansion and Doppler) and coolant heating 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, reactor coolant loops, and reactor point kinetics physical models will generally run orders of magnitude faster than real time on modern computing platforms.
  7. Unusual Features of the Program:
    The physical models in SASSYS-1 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 SASSYS-1 are shared with the SAS4A computer code. The SASSYS-1 computer code has been used as the computational engine for the EBR-II reactor plant simulator.
  9. Status:
    SASSYS-1 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.
  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 17-21, American Nuclear Society, 1994.
    2. P. L. Garner et al., "Development of a Graphical User Interface Allowing Use of the SASSYS-1 LMR Systems Analysis Code as an EBR-II Interactive Simulator", Proceedings of International Topical Meeting on Advanced Reactors Safety, Pittsburgh, PA, April 17-21, American Nuclear Society, 1994.
    3. 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 12-16, 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:
    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 SASSYS-1 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

 

Multimedia

The following videos are available:

  • Protected Loss of Flow Transient Simulation"Protected Loss of Flow Transient Simulation"
    Choose the video optimized for the bandwidth associated with your type of Internet connection; all videos are in Quicktime format.
    Low Bandwidth - [131 kbps, 2MB]
    Mid Bandwidth - [552 kbps, 10MB]
    High Bandwidth - [1448 kbps, 26MB]

Last Modified: Thu, July 18, 2013 10:48 AM

 

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