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CCC-0774 GRSAC.

GRSAC, Graphite Reactor Severe Accident Code

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1. NAME OR DESIGNATION OF PROGRAM:  GRSAC.
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2. COMPUTERS
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Program name Package id Status Status date
GRSAC CCC-0774/01 Arrived 23-SEP-2010

Machines used:

Package ID Orig. computer Test computer
CCC-0774/01 PC Windows
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3. DESCRIPTION OF PROGRAM OR FUNCTION

Graphite Reactor Severe Accident Code (GRSAC) is an interactive, personal computer-based simulation code for studying postulated severe accidents in gas-cooled reactors. Code features include user-generated input applicable to a variety of modular high-temperature gas-cooled reactor designs and gas-cooled test reactors.  Features include a ?smart front end? for detecting out-of-range or inconsistent data, online and off-line plotting, online help and documentation, and fast-running computational speeds. GRSAC model features include a three-dimensional thermal-fluid representation of the core, point kinetics for anticipated transients without scram (ATWS), and graphite oxidation for air ingress accident scenarios.
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4. METHODS

GRSAC was developed primarily to study a wide spectrum of core transient and heatup accident scenarios. It includes a detailed (~3000 nodes) 3-D hexagonal geometry T/F model for the core, plus T/F models for the RPV, SCS, and RCCS. The core T/F model consists of 163 radial nodes by 10 axial nodes for the fueled section (including the center reflector, if applicable). Axial coolant flows for the 163 channels are calculated independently; however, radial flows (which would occur more prominently in PBRs), are not accounted for. There are an additional 96 radial nodes for the side reflector, and two layers of axial nodes each for the top and bottom reflectors. There is an option to include neutronics (point kinetics), with xenon and samarium poisoning, to study accidents involving an ATWS. GRSAC also models air ingress accidents, simulating the oxidation of graphite core materials. The 3-D hexagonal geometry core thermal model allows for investigations of azimuthal temperature asymmetries in addition to axial and radial profiles. Variable core thermal properties are computed functions of temperature, and for prismatic core designs may also be dependent on block orientation and radiation damage. The annealing model for graphite can account for the increase in thermal conductivity occurring during LOFC accidents, which can have a significant effect on the predicted consequences. The primary coolant flow models cover the full ranges expected in both normal operation and accidents, including pressurized and depressurized accidents (and in between), for forced and natural circulation, for upward and downward flow, and for turbulent, laminar, and transition flow regimes. The primary loop pressure calculation can consider variable inventory (due to depressurization actions) and loop temperature changes, and use a simplified model for BOP temperatures and primary system pressure responses. The models for the RPV and the shield or RCCS are typically different for each of the various basic reactor models. FP release (for metal fuel) and Wigner stored-energy release models for graphite in the older model low-temperature gas reactors are also available.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The basic reactor designs that can be simulated using GRSAC, which the user may modify via the interface to a large (but limited) extent, are for various modular HTGR designs with particular adaptations for the PBMR (South Africa), the GT-MHR [Plutonium burner (U.S., Russia)], and the commercial uranium fueled designs such as the NGNP and VHTR (U.S.). Other code and model upgrades have focused on the operating test reactors such as the HTTR (Japan) and HTR-10 (China), and the recent HTR-PM power module design (China). Some of the features described are not functional for the older plant models.
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6. TYPICAL RUNNING TIME

Not directly stated but on the order of 10s of seconds.  ~25,000 times faster than real time for most accidents.
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9. STATUS
Package ID Status date Status
CCC-0774/01 23-SEP-2010 Masterfiled Arrived
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10. REFERENCES

- S. J. Ball, GRSAC User Manual, Oak Ridge National Laboratory report ORNL/TM-13697 (February 1999).
CCC-0774/01, included references:
- S. J. Ball:
A Graphite Reactor Severe Accident Code (GRSAC) for Modular High-Temperature
Gas-Cooled Reactors (HTGRs) User Manual, Oak Ridge National Laboratory report
ORNL/TM-2010/096 (June 2010).
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11. HARDWARE REQUIREMENTS:  An X86 processor.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
CCC-0774/01 C++
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13. SOFTWARE REQUIREMENTS:  Windows and Java (Either RTE or SDK).
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by:
                Radiation Shielding Information Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.

Developed by:   Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.
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16. MATERIAL AVAILABLE
CCC-0774/01
Readme file
Executables
Data files
Electronic documentation
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: accidents, containment, nuclear models, reactor safety.