Analysis of Semiscale Test S–LH–1 Using RELAP5/MOD2 (NUREG/IA–0064, GD/PE–N/725, PWR/HTWG/P(88)629(Rev))

On this page:

Download complete document

Publication Information

Date Published: April 1992

Prepared by:
P. C. Hall, D. R. Bull

National Power Nuclear
Barnett Way
Barnwood, Gloucester GL4 7RS
United Kingdom

Prepared as part of:
The Agreement on Research Participation and Technical Exchange
under the International Thermal-Hydraulic Code Assessment
and Application Program (ICAP)

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Availability Notice


The RELAP5/MOD2 code is being used by GDCD for calculating Small Break Loss of Coolant Accidents (SBLOCA) and pressurised transient sequences for the Sizewell 'B' PWR. These calculations are being carried out at the request of the Sizewell 'B' Project Management Team.

To assist in validating RELAP5/MOD2 for the above application, the code is being used by GDCD to model a number of small LOCA and pressurised fault simulation experiments carried out in various integral test facilities. The present report describes a RELAP5/MOD2 analysis of the small LOCA test S-LH-1 which was performed on the Semiscale Mod-2C facility. S-LH-1 simulated a small LOCA caused by a break in the cold leg pipework of an area equal to 5% of the cold leg flow area.

RELAP5/MOD2 gave reasonably accurate predictions of system thermal hydraulic behaviour. In particular, the hydrostatic core level depression resulting from the hold-up of water in the steam generator tubes and pump suction legs was well predicted. A reasonable prediction of core inventory was also obtained in the period of the test in which the core level fell as a result of coolant boil-off.

The code did not give an accurate prediction of the liquid distribution within the core during the uncovering phases. Consequently the fuel temperature excursions due to uncovery were not captured by the code. Failure to calculate the correct void fraction distribution and dryout behaviour is believed to be due to numerical approximations in representing the core by a small number of nodes, rather than due to errors in the physical models and correlations used in the code.

Based on the present study it is suggested that, in reactor analyses in which the potential for core uncovery occurs, the mixture level trajectory and peak fuel temperatures are calculated outside RELAP5 using a code employing a fine axial mesh, using boundary conditions from the RELAP5 analysis.

February 1989

Page Last Reviewed/Updated Tuesday, March 09, 2021