NUREG/IA-0106, TEC/L/0471/R91, "International Agreement Report, Assessment of PWR Steam Generator Modelling in RELAP5/MOD2."

NOTICE This report was prepared under an international cooperative agreement for the exchange of technical information. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use by such third party Would not infringe privately owned rights. The United Kingdom has consented to the publication of this report as a USNRC document in order to allow the widest possible circulation among the reactor safety community. Neither the United States Government nor the United Kingdom or any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability of responsibility for any third party's use, or the results of such use, or any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. SUMMARY An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed-including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3. Conclusions 1. RELAP5/MOD2 under-predicts SG heat transfer under steady-state normal operating and start-up conditions. If the code is initialised with the correct primary side conditions, this is reflected by an under-prediction of the secondary side pressure. The effect is seen in both plant and rig calculations, although it tends to reduce as both reactor power and scale reduce. For the Sizewell 'B' SG operating at full load conditions, the error in secondary side pressure is around 3.5 bar. The deficiency can be attributed to the application of the Chen correlation, which was developed using data from flows in tubes and annuli, to calculate the boiling heat transfer coefficient on the secondary side of the U-tube bundle. Although …