A first look at LOCAs in the SBWR using RELAP5/MOD3 [electronic resource].
- Published:
- Rockville, Md. : U.S. Nuclear Regulatory Commission, 1992.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy. - Physical Description:
- Pages: (17 pages) : digital, PDF file
- Additional Creators:
- U.S. Nuclear Regulatory Commission and United States. Department of Energy. Office of Scientific and Technical Information
Access Online
- Restrictions on Access:
- Free-to-read Unrestricted online access
- Summary:
- The General Electric Company (GE) is designing an advanced light-water reactor, the Simplified Boiling Water Reactor (SBWR), that utilizes passive safety concepts. The SBWR reactor coolant system will operate on natural circulation with decay heat removal and emergency core coolant injection being provided by passive, gravity-driven systems. The Idaho National Engineering Laboratory has developed an input model of the SBWR for the RELAP5/MOD3 thermal-hydraulic safety analysis code. Preliminary calculations have been performed to simulate three loss-of-coolant accidents: (1) a main steam line break, (2) spurious opening of one automatic depressurization valve, and (3) the rupture of the bottom drain line. Results from these three calculations were, in general, intuitively reasonable. The analyses revealed that the input model, which was created with preliminary design data, needs to be updated to reflect the current SBWR design. Nodalization of certain regions will also need to be improved. The results of the main steam line break calculation were compared to a similar TRACG calculation presented in GE's Standard Safety Analysis Report. Comparisons of the preliminary RELAP5/MOD3 results to TRACG results indicated good qualitative agreement.
- Report Numbers:
- E 1.99:egg-m-92511
E 1.99: conf-9210204--4
conf-9210204--4
egg-m-92511 - Subject(s):
- Other Subject(s):
- Bwr Type Reactors
- Loss Of Coolant
- Computer Calculations
- Computerized Simulation
- Heat Transfer
- Hydraulics
- R Codes
- Reactor Cooling Systems
- Reactor Safety
- T Codes
- Accidents
- Computer Codes
- Cooling Systems
- Energy Transfer
- Enriched Uranium Reactors
- Fluid Mechanics
- Mechanics
- Power Reactors
- Reactor Accidents
- Reactor Components
- Reactors
- Safety
- Simulation
- Thermal Reactors
- Water Cooled Reactors
- Water Moderated Reactors
- Note:
- Published through SciTech Connect.
01/01/1992.
"egg-m-92511"
" conf-9210204--4"
"DE93005252"
Water reactor safety information meeting, Washington, DC (United States), 21-23 Oct 1992.
Shaw, R.A.; Ghan, L.S.; Kullberg, C.M.
EG and G Idaho, Inc., Idaho Falls, ID (United States) - Funding Information:
- AC07-76ID01570
View MARC record | catkey: 14367021