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Analysis of Semiscale Test S–LH–2 Using RELAP5/MOD2 (NUREG/IA–0065, GD/PE–N/745, PWR/HTWG/P(89)708)

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Publication Information

Date Published: April 1992

Prepared by:
P. Brodie, P. C. Hall

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

Summary

The RELAP5/MOD2 code is being used by National Power Nuclear Technology-Division 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 and the Health and Safety Department.

To assist in validating RELAP5/MOD2 for the above application, the code is being used to model a number of small LOCA and pressurised fault simulation experiments carried out in integral test facilities. The present report describes a RELAP5/MOD2 analysis of the small LOCA test S-LH-2 which was performed on the Semiscale Mod-2C Facility. S-LH-2 simulated a SBLOCA caused by a break in the cold leg pipework of an area equal to 5% of the cold leg flow area. S-LH-2 was identical to the earlier test S-LH-1 except for an increase in the flow area between the upper plenum and the cold leg which resulted in the core bypass flow increasing from 0.9% to 3.0%.

RELAP5/MOD2 gave reasonably accurate predictions of system thermal hydraulic behaviour but failed to calculate the core dryout which occurred due to coolant boil-off prior to accumulator injection. The error is believed due to combinations of errors in calculating the liquid inventory in the core and steam generators, and incorrect modelling of the void fraction gradient within the core.

Task No. G212

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