Actions for Development of RELAP
Development of RELAP/SLIP for the semiscale blowdown heat transfer test S-02-6 (NRC Standard Problem 6).
- Author
- Lyczkowski, R. W.
- Published
- United States : [publisher not identified], 1976.
[Oak Ridge, Tennessee] : [U.S. Atomic Energy Commission], 1976. - Physical Description
- microfiche : negative ; 11 x 15 cm
- Additional Creators
- Fujita, N., Irani, A. A., Mecham, D. C., Moore, K. V., and Sawtelle, G. R.
- Summary
- The RELAP4/SLIP code was developed in order to provide a realistic prediction of NRC Standard Problem 6 (Test S-02-6). RELAP/SLIP is a modified version of RELAP4-95 and includes a first level effort, state-of-the-art dynamic slip calculation in all vertical and horizontal junctions to account for unequal phase velocity and phase separation. The dynamic slip model accounts for transient inertial effects of each phase. This report contains the derivation of the dynamic slip field equations, a description of the modifications to the RELAP4 numerical scheme and recommendations for improved code performance. Several analytical problems and some experiment simulations including NRC Standard Problems 1 and 5 were analyzed in an effort to check out the RELAP/SLIP code. RELAP/SLIP calculations compared favorably with homogeneous equilibrium experimental data and previous RELAP4-95 predictions. To the best of the authors' knowledge, the consistent, complete and realistic computation of unequal phase velocities has been achieved for the first time in any general systems code with the successful simulation of the simple problems and Standard Problem 5 using RELAP/SLIP. This milestone is thus a signal achievement in the advancement of the state-of-the-art of transient two-phase, two-velocity, compressible flow analysis.
- Report Numbers
- EPRI-NP-343
- Other Subject(s)
- 21 specific nuclear reactors and associated plants
- 210100 - power reactors, nonbreeding, light-water moderated, boiling water cooled
- 210200 - power reactors, nonbreeding, light-water moderated, nonboiling water cooled
- 22 general studies of nuclear reactors
- 220900 - nuclear reactor technology- reactor safety
- Accidents
- Blowdown
- Bwr type reactors
- Computer codes
- Fluid flow
- Fluid mechanics
- Hydraulics
- Loss of coolant
- Mechanics
- Pwr type reactors
- R codes
- Reactor accidents
- Reactors
- Simulation
- Two-phase flow
- Water cooled reactors
- Water moderated reactors
- Collection
- U.S. Atomic Energy Commission depository collection.
- Note
- OSTI Identifier 7253453
Research organization: Energy, Inc., Idaho Falls, ID (USA).
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