Published 1990 | Version v1
Report Open

Assessment of RAMONA-3B methodology with FRIGG dynamic tests

Description

The computer codes used at Brookhaven National Laboratory to compute BWR safety parameters are the Engineering Plant Analyzer (EPA) and RAMONA-3B/MOD1. Both codes have the same methodology for modeling thermal hydraulic phenomena: drift-flux formulation, two-phase multipliers for the wall friction and form losses calculations, and the momentum integral approach for spatial integration of the loop momentum equations. Both codes use explicit integration methods for solving ordinary differential equations. It is concluded that both the codes are capable of modelling the instability problems for a BWR. The accuracy of thermohydraulics codes predictions was assessed by modelling oscillatory FRIGG tests. Nodalizations studies showed that 24 axial nodes were sufficient for a converged solution, 12 axial nodes produced an error of 4.4% in the gain of the power to flow transfer function. The code predicted consistently the effects of power and inlet subcooling on gain and system resonance frequency. The comparisons showed that the code predicted the peak gains with a mean difference from experiments of 7% ± 30% for all the tests modeled. The uncertainty in the experimental data is -11% to +12%. The mean difference in the predicted frequency at the peak gain is -6% ± 14%

Availability note (English)

MF available from INIS under the Report Number; OSTI as DE92011487; NTIS; INIS; US Govt. Printing Office Dep.

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Additional details

Publishing Information

Imprint Pagination
18 p.
Report number
BNL-NUREG--45315

Conference

Title
International workshop on boiling water reactor stability.
Dates
17-19 Oct 1990.
Place
Upton, NY (United States).

Optional Information

Contract/Grant/Project number
Contract AC02-76CH00016
Funding organization
Nuclear Regulatory Commission, Washington, DC (United States).
Secondary number(s)
CONF-9010224--6.